Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

NASA Astrophysics Data System (ADS)

Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

2017-01-01

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance duringmore » LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.« less

40 CFR 59.506 - How do I demonstrate compliance if I manufacture multi-component kits?

Code of Federal Regulations, 2011 CFR

2011-07-01

... multi-component kits as defined in § 59.503, then the Kit PWR must not exceed the Total Reactivity Limit. (b) You must calculate the Kit PWR and the Total Reactivity Limit as follows: (1) KIT PWR = (PWR(1) × W1) + (PWR(2) × W2) +. ...+ (PWR(n) × Wn) (2) Total Reactivity Limit = (RL1 × W1) + (RL2 × W2...

40 CFR 59.506 - How do I demonstrate compliance if I manufacture multi-component kits?

Code of Federal Regulations, 2014 CFR

2014-07-01

... multi-component kits as defined in § 59.503, then the Kit PWR must not exceed the Total Reactivity Limit. (b) You must calculate the Kit PWR and the Total Reactivity Limit as follows: (1) KIT PWR = (PWR(1) × W1) + (PWR(2) × W2) +. ...+ (PWR(n) × Wn) (2) Total Reactivity Limit = (RL1 × W1) + (RL2 × W2...

40 CFR 59.506 - How do I demonstrate compliance if I manufacture multi-component kits?

Code of Federal Regulations, 2012 CFR

2012-07-01

... multi-component kits as defined in § 59.503, then the Kit PWR must not exceed the Total Reactivity Limit. (b) You must calculate the Kit PWR and the Total Reactivity Limit as follows: (1) KIT PWR = (PWR(1) × W1) + (PWR(2) × W2) +. ...+ (PWR(n) × Wn) (2) Total Reactivity Limit = (RL1 × W1) + (RL2 × W2...

40 CFR 59.506 - How do I demonstrate compliance if I manufacture multi-component kits?

Code of Federal Regulations, 2013 CFR

2013-07-01

... multi-component kits as defined in § 59.503, then the Kit PWR must not exceed the Total Reactivity Limit. (b) You must calculate the Kit PWR and the Total Reactivity Limit as follows: (1) KIT PWR = (PWR(1) × W1) + (PWR(2) × W2) +. ...+ (PWR(n) × Wn) (2) Total Reactivity Limit = (RL1 × W1) + (RL2 × W2...

Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

NASA Astrophysics Data System (ADS)

Burdo, James S.

This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the difference was less than 1.00, allowing for the existence of the difference within the margin of error. The second was whether the difference between the two values was big enough to prevent their error bars from overlapping. Error analysis was performed both using a one second count and pseudo-Maxwell statistics for a projected 60 second count, giving four criteria for detection. The number of guide tubes meeting these criteria was compared and graphed for each case. Further analysis at extremes of high and low enrichment and long and short burnup times was done using data from assemblies at the Beaver Valley 1 and 2 PWR. In all neutron flux cases, at least two guide tube locations meet all the criteria for detection of pin diversion. At least one location in almost all of the gamma flux cases does. These results show that placing detectors in the empty guide tubes of spent fuel bundles to identify possible pin diversion is feasible.

Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal,more » 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.

Severe accident management can be defined as the use of existing and/or altemative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there may be several options available to the operator, and each involves phenomenological and operational considerations regarding uncertainty. Operational uncertainties include operator, system and instrumentation behavior during an accident. A framework based on decision trees and influence diagrams has been developed which incorporates such criteria as feasibility, effectiveness, and adverse effects, for evaluating potential severe accident management strategies. The framework is also capable ofmore » propagating both data and model uncertainty. It is applied to several potential strategies including PWR cavity flooding, BWR drywell flooding, PWR depressurization and PWR feed and bleed.« less

PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew Ellis; Derek Gaston; Benoit Forget

In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes.more » An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.« less

Community concepts of poverty: an application to premium exemptions in Ghana’s National Health Insurance Scheme

PubMed Central

2013-01-01

Background Poverty is multi dimensional. Beyond the quantitative and tangible issues related to inadequate income it also has equally important social, more intangible and difficult if not impossible to quantify dimensions. In 2009, we explored these social and relativist dimension of poverty in five communities in the South of Ghana with differing socio economic characteristics to inform the development and implementation of policies and programs to identify and target the poor for premium exemptions under Ghana’s National Health Insurance Scheme. Methods We employed participatory wealth ranking (PWR) a qualitative tool for the exploration of community concepts, identification and ranking of households into socioeconomic groups. Key informants within the community ranked households into wealth categories after discussing in detail concepts and indicators of poverty. Results Community defined indicators of poverty covered themes related to type of employment, educational attainment of children, food availability, physical appearance, housing conditions, asset ownership, health seeking behavior, social exclusion and marginalization. The poverty indicators discussed shared commonalities but contrasted in the patterns of ranking per community. Conclusion The in-depth nature of the PWR process precludes it from being used for identification of the poor on a large national scale in a program such as the NHIS. However, PWR can provide valuable qualitative input to enrich discussions, development and implementation of policies, programs and tools for large scale interventions and targeting of the poor for social welfare programs such as premium exemption for health care. PMID:23497484

Community concepts of poverty: an application to premium exemptions in Ghana's National Health Insurance Scheme.

PubMed

Aryeetey, Genevieve C; Jehu-Appiah, Caroline; Kotoh, Agnes M; Spaan, Ernst; Arhinful, Daniel K; Baltussen, Rob; van der Geest, Sjaak; Agyepong, Irene A

2013-03-14

Poverty is multi dimensional. Beyond the quantitative and tangible issues related to inadequate income it also has equally important social, more intangible and difficult if not impossible to quantify dimensions. In 2009, we explored these social and relativist dimension of poverty in five communities in the South of Ghana with differing socio economic characteristics to inform the development and implementation of policies and programs to identify and target the poor for premium exemptions under Ghana's National Health Insurance Scheme. We employed participatory wealth ranking (PWR) a qualitative tool for the exploration of community concepts, identification and ranking of households into socioeconomic groups. Key informants within the community ranked households into wealth categories after discussing in detail concepts and indicators of poverty. Community defined indicators of poverty covered themes related to type of employment, educational attainment of children, food availability, physical appearance, housing conditions, asset ownership, health seeking behavior, social exclusion and marginalization. The poverty indicators discussed shared commonalities but contrasted in the patterns of ranking per community. The in-depth nature of the PWR process precludes it from being used for identification of the poor on a large national scale in a program such as the NHIS. However, PWR can provide valuable qualitative input to enrich discussions, development and implementation of policies, programs and tools for large scale interventions and targeting of the poor for social welfare programs such as premium exemption for health care.

77 FR 37795 - Airworthiness Directives; Dassault Aviation Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-25

... display of ELEC:LH ESS PWR LO or ELEC:LH ESS NO PWR (Abnormal procedure 3-190-40), land at nearest suitable airport Upon display of ELEC:RH ESS PWR LO and ELEC:RH ESS NO PWR (Abnormal procedure 3-190-45...

The influence of psychological factors on post-partum weight retention at 9 months.

PubMed

Phillips, Joanne; King, Ross; Skouteris, Helen

2014-11-01

Post-partum weight retention (PWR) has been identified as a critical pathway for long-term overweight and obesity. In recent years, psychological factors have been demonstrated to play a key role in contributing to and maintaining PWR. Therefore, the aim of this study was to explore the relationship between post-partum psychological distress and PWR at 9 months, after controlling for maternal weight factors, sleep quality, sociocontextual influences, and maternal behaviours. Pregnant women (N = 126) completed a series of questionnaires at multiple time points from early pregnancy until 9 months post-partum. Hierarchical regression indicated that gestational weight gain, shorter duration (6 months or less) of breastfeeding, and post-partum body dissatisfaction at 3 and 6 months are associated with higher PWR at 9 months; stress, depression, and anxiety had minimal influence. Interventions aimed at preventing excessive PWR should specifically target the prevention of body dissatisfaction and excessive weight gain during pregnancy. What is already known on this subject? Post-partum weight retention (PWR) is a critical pathway for long-term overweight and obesity. Causes of PWR are complex and multifactorial. There is increasing evidence that psychological factors play a key role in predicting high PWR. What does this study add? Post-partum body dissatisfaction at 3 and 6 months is associated with PWR at 9 months post-birth. Post-partum depression, stress and anxiety have less influence on PWR at 9 months. Interventions aimed at preventing excessive PWR should target body dissatisfaction. © 2013 The British Psychological Society.

40 CFR Appendix A to Part 76 - Phase I Affected Coal-Fired Utility Units With Group 1 or Cell Burner Boilers

Code of Federal Regulations, 2012 CFR

2012-07-01

... MONTROSE 2 KANSAS CITY PWR & LT. MISSOURI MONTROSE 3 KANSAS CITY PWR & LT. NEW YORK DUNKIRK 3 NIAGARA MOHAWK PWR. NEW YORK DUNKIRK 4 NIAGARA MOHAWK PWR. NEW YORK GREENIDGE 6 NY STATE ELEC & GAS. NEW YORK...

40 CFR Appendix A to Part 76 - Phase I Affected Coal-Fired Utility Units With Group 1 or Cell Burner Boilers

Code of Federal Regulations, 2014 CFR

2014-07-01

... MONTROSE 2 KANSAS CITY PWR & LT. MISSOURI MONTROSE 3 KANSAS CITY PWR & LT. NEW YORK DUNKIRK 3 NIAGARA MOHAWK PWR. NEW YORK DUNKIRK 4 NIAGARA MOHAWK PWR. NEW YORK GREENIDGE 6 NY STATE ELEC & GAS. NEW YORK...

40 CFR Appendix A to Part 76 - Phase I Affected Coal-Fired Utility Units With Group 1 or Cell Burner Boilers

Code of Federal Regulations, 2013 CFR

2013-07-01

... MONTROSE 2 KANSAS CITY PWR & LT. MISSOURI MONTROSE 3 KANSAS CITY PWR & LT. NEW YORK DUNKIRK 3 NIAGARA MOHAWK PWR. NEW YORK DUNKIRK 4 NIAGARA MOHAWK PWR. NEW YORK GREENIDGE 6 NY STATE ELEC & GAS. NEW YORK...

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

VERA Core Simulator Methodology for PWR Cycle Depletion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel

2015-01-01

This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclearmore » reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.« less

Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

NASA Technical Reports Server (NTRS)

Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

1986-01-01

The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

Extension of the Bgl Broad Group Cross Section Library

NASA Astrophysics Data System (ADS)

Kirilova, Desislava; Belousov, Sergey; Ilieva, Krassimira

2009-08-01

The broad group cross-section libraries BUGLE and BGL are applied for reactor shielding calculation using the DOORS package based on discrete ordinates method and multigroup approximation of the neutron cross-sections. BUGLE and BGL libraries are problem oriented for PWR or VVER type of reactors respectively. They had been generated by collapsing the problem independent fine group library VITAMIN-B6 applying PWR and VVER one-dimensional radial model of the reactor middle plane using the SCALE software package. The surveillance assemblies (SA) of VVER-1000/320 are located on the baffle above the reactor core upper edge in a region where geometry and materials differ from those of the middle plane and the neutron field gradient is very high which would result in a different neutron spectrum. That is why the application of the fore-mentioned libraries for the neutron fluence calculation in the region of SA could lead to an additional inaccuracy. This was the main reason to study the necessity for an extension of the BGL library with cross-sections appropriate for the SA region. Comparative analysis of the neutron spectra of the SA region calculated by the VITAMIN-B6 and BGL libraries using the two-dimensional code DORT have been done with purpose to evaluate the BGL applicability for SA calculation.

Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feldman, E. M.; Patton, Jr., M. L.; Sackett, K. E.

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulatemore » emergency core coolant injection in a PWR, with the flow rate based on system volume scaling.« less

Warm prestress effects in fracture-margin assessment of PWR-RPVs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shum, D.K.M.

This paper examines various issues that would impact the incorporation of warm prestress (WPS) effects in the fracture-margin assessment of reactor pressure vessels (RPVs). By way of an example problem, possible beneficial effects of including type-I WPS in the assessment of an RPV subjected to a small break loss of coolant accident are described. In addition, the need to consider possible loss of constraint effects when interpreting available small specimen WPS-enhanced fracture toughness data is demonstrated through two- and three-dimensional local crack-lip field analyses of a compact tension specimen. Finally, a hybrid correlative-predictive model of WPS base on J-Q theorymore » and the Ritchie-Knott-Rice model is applied to a small scale yielding boundary layer formulation to investigate near crack-tip fields under varying degrees of loading and unloading.« less

Modeling local chemistry in PWR steam generator crevices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Millett, P.J.

1997-02-01

Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledgemore » of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.« less

Plasmid partition system of the P1par family from the pWR100 virulence plasmid of Shigella flexneri.

PubMed

Sergueev, Kirill; Dabrazhynetskaya, Alena; Austin, Stuart

2005-05-01

P1par family members promote the active segregation of a variety of plasmids and plasmid prophages in gram-negative bacteria. Each has genes for ParA and ParB proteins, followed by a parS partition site. The large virulence plasmid pWR100 of Shigella flexneri contains a new P1par family member: pWR100par. Although typical parA and parB genes are present, the putative pWR100parS site is atypical in sequence and organization. However, pWR100parS promoted accurate plasmid partition in Escherichia coli when the pWR100 Par proteins were supplied. Unique BoxB hexamer motifs within parS define species specificities among previously described family members. Although substantially different from P1parS from the P1 plasmid prophage of E. coli, pWR100parS has the same BoxB sequence. As predicted, the species specificity of the two types proved identical. They also shared partition-mediated incompatibility, consistent with the proposed mechanistic link between incompatibility and species specificity. Among several informative sequence differences between pWR100parS and P1parS is the presence of a 21-bp insert at the center of the pWR100parS site. Deletion of this insert left much of the parS activity intact. Tolerance of central inserts with integral numbers of helical DNA turns reflects the critical topology of these sites, which are bent by binding the host IHF protein.

77 FR 15293 - Airworthiness Directives; Dassault Aviation Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-15

...-190-20), land at nearest suitable airport Upon display of ELEC:LH ESS PWR LO or ELEC:LH ESS NO PWR (Abnormal procedure 3-190-40), land at nearest suitable airport Upon display of ELEC:RH ESS PWR LO and ELEC...

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwon, Tae-Soon; Yun, Byong-Jo; Euh, Dong-Jin

Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focusedmore » on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.« less

Implementing a Nuclear Power Plant Model for Evaluating Load-Following Capability on a Small Grid

NASA Astrophysics Data System (ADS)

Arda, Samet Egemen

A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Adequacy of the isolated reactor core, the isolated steam generator, and the complete PWR models are tested in Matlab/Simulink and dynamic responses are compared with the test results obtained from the H. B. Robinson NPP. Test results illustrate that the developed models represents the dynamic features of real-physical systems and are capable of predicting responses due to small perturbations of external reactivity and steam valve opening. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a small power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The conclusion is that the NPP can respond to a disturbance in the power system without exceeding any design and safety limits if appropriate operational conditions, such as achieving the NPP turbine control by adjusting the speed of the steam valve, are met. In other words, the NPP can participate in the control of system frequency and improve the overall power system performance.

PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loftus, M J; Hochreiter, L E; McGuire, M F

This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

Analysis of the return to power scenario following a LBLOCA in a PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Macian, R.; Tyler, T.N.; Mahaffy, J.H.

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus,more » the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.« less

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ishii, Mamoru

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results andmore » models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup power density on the flow instability. The experimental startup transient results showed the existence of three different flow instability mechanisms, i.e., flashing instability, condensation induced flow instability, and density wave oscillations. In addition, the void-reactivity feedback did not have significant effects on the flow instability during the startup transients for NMR-50. ii Several initial startup procedures with different power ramp rates were experimentally investigated to eliminate the flow instabilities observed from the startup transients. Particularly, the very slow startup transient and pressurized startup transient tests were performed and compared. It was found that the very slow startup transients by applying very small power density can eliminate the flashing oscillations in the single-phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. The initially pressurized startup procedure was tested to eliminate the flashing instability during the startup transients as well. The pressurized startup procedure included the initial pressurization, heat-up, and venting process. The startup transient tests showed that the pressurized startup procedure could eliminate the flow instability during the transition from single-phase flow to two-phase flow at low pressure conditions. The experimental results indicated that both startup procedures were applicable to the initial startup of NMR. However, the pressurized startup procedures might be preferred due to short operating hours required. In order to have a deeper understanding of natural circulation flow instability, the quasi-steady tests were performed using the test facility installed with preheater and subcooler. The effect of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback were investigated in the quasi-steady state tests. The experimental stability boundaries were determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR-type SMR. Two kinds of experiments, normal blowdown event and cold blowdown event, were experimentally investigated and compared with code predictions. The normal blowdown event was experimentally simulated since an initial condition where the pressure was lower than the designed pressure of the experiment facility, while the code prediction of blowdown started from the normal operation condition. Important thermal hydraulic parameters including reactor pressure vessel (RPV) pressure, containment pressure, local void fraction and temperature, pressure drop and natural circulation flow rate were measured and analyzed during the blowdown event. The pressure and water level transients are similar to the experimental results published by NuScale [51], which proves the capability of current loop in simulating the thermal hydraulic transient of real PWR-type SMR. During the 20000s blowdown experiment, water level in the core was always above the active fuel assemble during the experiment and proved the safety of natural circulation cooling and water recycling design of PWR-type SMR. Besides, pressure, temperature, and water level transient can be accurately predicted by RELAP5 code. However, the oscillations of natural circulation flow rate, water level and pressure drops were observed during the blowdown transients. This kind of flow oscillations are related to the water level and the location upper plenum, which is a path for coolant flow from chimney to steam generator and down comer. In order to investigate the transients start from the opening of ADS valve in both experimental and numerical way, the cold blow-down experiment is conducted. For the cold blowdown event, different from setting both reactor iv pressure vessel (RPV) and containment at high temperature and pressure, only RPV was heated close to the highest designed pressure and then open the ADS valve, same process was predicted using RELAP5 code. By doing cold blowdown experiment, the entire transients from the opening of ADS can be investigated by code and benchmarked with experimental data. Similar flow instability observed in the cold blowdown experiment. The comparison between code prediction and experiment data showed that the RELAP5 code can successfully predict the pressure void fraction and temperature transient during the cold blowdown event with limited error, but numerical instability exists in predicting natural circulation flow rate. Besides, the code is lack of capability in predicting the water level related flow instability observed in experiments.« less

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

NASA Astrophysics Data System (ADS)

Cockeram, B. V.; Kammenzind, B. F.

2018-02-01

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

Logistics Modeling of Emplacement Rate and Duration of Operations for Generic Geologic Repository Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinina, Elena Arkadievna; Hardin, Ernest

This study identified potential geologic repository concepts for disposal of spent nuclear fuel (SNF) and (2) evaluated the achievable repository waste emplacement rate and the time required to complete the disposal for these concepts. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel. The results of this study provide an important input for the rough-order-of-magnitude (ROM) disposal cost analysis. The disposal concepts cover three major categories of host geologic media: crystalline or hard rock, salt, and argillaceous rock. Four waste package sizes are considered: 4PWR/9BWR; 12PWR/21BWR; 21PWR/44BWR, and dual purpose canisters (DPCs). The DPC concepts assumemore » that the existing canisters will be sealed into disposal overpacks for direct disposal. Each concept assumes one of the following emplacement power limits for either emplacement or repository closure: 1.7 kW; 2.2 kW; 5.5 kW; 10 kW; 11.5 kW, and 18 kW.« less

3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Barua, Bipul; Listwan, Joseph

In financial year 2017, we are focusing on developing a mechanistic fatigue model of surge line pipes for pressurized water reactors (PWRs). To that end, we plan to perform the following tasks: (1) conduct stress- and strain-controlled fatigue testing of surge-line base metal such as 316 stainless steel (SS) under constant, variable, and random fatigue loading, (2) develop cyclic plasticity material models of 316 SS, (3) develop one-dimensional (1D) analytical or closed-form model to validate the material models and to understand the mechanics associated with 316 SS cyclic hardening and/or softening, (4) develop three-dimensional (3D) finite element (FE) models withmore » implementation of evolutionary cyclic plasticity, and (5) develop computational fluid dynamics (CFD) model for thermal stratification, thermal-mechanical stress, and fatigue of example reactor components, such as a PWR surge line under plant heat-up, cool-down, and normal operation with/without grid-load-following. This semi-annual progress report presents the work completed on the above tasks for a 316 SS laboratory-scale specimen subjected to strain-controlled cyclic loading with constant, variable, and random amplitude. This is the first time that the accurate 3D-FE modeling of the specimen for its entire fatigue life, including the hardening and softening behavior, has been achieved. We anticipate that this work will pave the way for the development of a fully mechanistic-computer model that can be used for fatigue evaluation of safety-critical metallic components, which are traditionally evaluated by heavy reliance on time-consuming and costly test-based approaches. This basic research will not only help the nuclear reactor industry for fatigue evaluation of reactor components in a cost effective and less time-consuming way, but will also help other safety-related industries, such as aerospace, which is heavily dependent on test-based approaches, where a single full-scale fatigue test can cost millions of dollars and require years of effort to conduct. Toward our goal of demonstration of fully mechanistic fatigue evaluation of reactor components, we also started work on developing a component-level computer model of reactor components, such as 316 SS surge line pipe. This requires developing a thermal-mechanical stress analysis model of the reactor surge line, which, in turn, requires time-dependent temperature and stratification information along the boundary of the pipe. Toward that goal, CFD models of surge lines are being developed. In this report, we also present some preliminary results showing the temperature conditions along the surge line wall under reactor heat-up, cool-down, and steady-state power operation.« less

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Cuadra, Arantxa; Todosow, Michael

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order tomore » assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.« less

Trends in Global Vegetation Activity and Climatic Drivers Indicate a Decoupled Response to Climate Change.

PubMed

Schut, Antonius G T; Ivits, Eva; Conijn, Jacob G; Ten Brink, Ben; Fensholt, Rasmus

2015-01-01

Detailed understanding of a possible decoupling between climatic drivers of plant productivity and the response of ecosystems vegetation is required. We compared trends in six NDVI metrics (1982-2010) derived from the GIMMS3g dataset with modelled biomass productivity and assessed uncertainty in trend estimates. Annual total biomass weight (TBW) was calculated with the LINPAC model. Trends were determined using a simple linear regression, a Thiel-Sen medium slope and a piecewise regression (PWR) with two segments. Values of NDVI metrics were related to Net Primary Production (MODIS-NPP) and TBW per biome and land-use type. The simple linear and Thiel-Sen trends did not differ much whereas PWR increased the fraction of explained variation, depending on the NDVI metric considered. A positive trend in TBW indicating more favorable climatic conditions was found for 24% of pixels on land, and for 5% a negative trend. A decoupled trend, indicating positive TBW trends and monotonic negative or segmented and negative NDVI trends, was observed for 17-36% of all productive areas depending on the NDVI metric used. For only 1-2% of all pixels in productive areas, a diverging and greening trend was found despite a strong negative trend in TBW. The choice of NDVI metric used strongly affected outcomes on regional scales and differences in the fraction of explained variation in MODIS-NPP between biomes were large, and a combination of NDVI metrics is recommended for global studies. We have found an increasing difference between trends in climatic drivers and observed NDVI for large parts of the globe. Our findings suggest that future scenarios must consider impacts of constraints on plant growth such as extremes in weather and nutrient availability to predict changes in NPP and CO2 sequestration capacity.

Sub-Scale Testing and Development of the J-2X Fuel Turbopump Inducer

NASA Technical Reports Server (NTRS)

Sargent, Scott R.; Becht, David G.

2011-01-01

In the early stages of the J-2X upper stage engine program, various inducer configurations proposed for use in the fuel turbopump (FTP) were tested in water. The primary objectives of this test effort were twofold. First, to obtain a more comprehensive data set than that which existed in the Pratt & Whitney Rocketdyne (PWR) historical archives from the original J-2S program, and second, to supplement that data set with information regarding the cavitation induced vibrations for both the historical J-2S configuration as well as those tested for the J-2X program. The J-2X FTP inducer, which actually consists of an inducer stage mechanically attached to a kicker stage, underwent 4 primary iterations utilizing sub-scaled test articles manufactured and tested in PWR's Engineering Development Laboratory (EDL). The kicker remained unchanged throughout the test series. The four inducer configurations tested retained many of the basic design features of the J-2S inducer, but also included variations on leading edge blade thickness and blade angle distribution, primarily aimed at improving suction performance at higher flow coefficients. From these data sets, the effects of the tested design variables on hydrodynamic performance and cavitation instabilities were discerned. A limited comparison of impact to the inducer efficiency was determined as well.

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

NASA Astrophysics Data System (ADS)

Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

2018-02-01

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

The fractalline properties of experimentally simulated PWR fuel crud

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

2018-02-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

Assessment for advanced fuel cycle options in CANDU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Noninvasive and Real-Time Plasmon Waveguide Resonance Thermometry

PubMed Central

Zhang, Pengfei; Liu, Le; He, Yonghong; Zhou, Yanfei; Ji, Yanhong; Ma, Hui

2015-01-01

In this paper, the noninvasive and real-time plasmon waveguide resonance (PWR) thermometry is reported theoretically and demonstrated experimentally. Owing to the enhanced evanescent field and thermal shield effect of its dielectric layer, a PWR thermometer permits accurate temperature sensing and has a wide dynamic range. A temperature measurement sensitivity of 9.4 × 10−3 °C is achieved and the thermo optic coefficient nonlinearity is measured in the experiment. The measurement of water cooling processes distributed in one dimension reveals that a PWR thermometer allows real-time temperature sensing and has potential to be applied for thermal gradient analysis. Apart from this, the PWR thermometer has the advantages of low cost and simple structure, since our transduction scheme can be constructed with conventional optical components and commercial coating techniques. PMID:25871718

Structural Integrity of Water Reactor Pressure Boundary Components.

DTIC Science & Technology

1981-02-20

environment, and load waveform parameters . A theory of the influence of dissolved oxygen content on the fatigue crack growth results in simulated PWR ...simulated PWR coolant is - (Continues ) DD IJN7 1473 EDITION OF I NOV S ..OSL- -C 2 S/ 0102-LF-014-6601 S1ECURITY CLASSI1FICATION OF THIS PAGE (When...not seem to influence the data, which was produced for a load ratio of 0.2 and a simulated PWR coolant environment. Test results for A106 Grade C piping

J-2X Turbopump Cavitation Diagnostics

NASA Technical Reports Server (NTRS)

Santi, I. Michael; Butas, John P.; Tyler, Thomas R., Jr.; Aguilar, Robert; Sowers, T. Shane

2010-01-01

The J-2X is the upper stage engine currently being designed by Pratt & Whitney Rocketdyne (PWR) for the Ares I Crew Launch Vehicle (CLV). Propellant supply requirements for the J-2X are defined by the Ares Upper Stage to J-2X Interface Control Document (ICD). Supply conditions outside ICD defined start or run boxes can induce turbopump cavitation leading to interruption of J-2X propellant flow during hot fire operation. In severe cases, cavitation can lead to uncontained engine failure with the potential to cause a vehicle catastrophic event. Turbopump and engine system performance models supported by system design information and test data are required to predict existence, severity, and consequences of a cavitation event. A cavitation model for each of the J-2X fuel and oxidizer turbopumps was developed using data from pump water flow test facilities at Pratt & Whitney Rocketdyne (PWR) and Marshall Space Flight Center (MSFC) together with data from Powerpack 1A testing at Stennis Space Center (SSC) and from heritage systems. These component models were implemented within the PWR J-2X Real Time Model (RTM) to provide a foundation for predicting system level effects following turbopump cavitation. The RTM serves as a general failure simulation platform supporting estimation of J-2X redline system effectiveness. A study to compare cavitation induced conditions with component level structural limit thresholds throughout the engine was performed using the RTM. Results provided insight into system level turbopump cavitation effects and redline system effectiveness in preventing structural limit violations. A need to better understand structural limits and redline system failure mitigation potential in the event of fuel side cavitation was indicated. This paper examines study results, efforts to mature J-2X turbopump cavitation models and structural limits, and issues with engine redline detection of cavitation and the use of vehicle-side abort triggers to augment the engine redline system.

Evaluation of CASL boiling model for DNB performance in full scale 5x5 fuel bundle with spacer grids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Seung Jun

As one of main tasks for FY17 CASL-THM activity, Evaluation study on applicability of the CASL baseline boiling model for 5x5 DNB application is conducted and the predictive capability of the DNB analysis is reported here. While the baseline CASL-boiling model (GEN- 1A) approach has been successfully implemented and validated with a single pipe application in the previous year’s task, the extended DNB validation for realistic sub-channels with detailed spacer grid configurations are tasked in FY17. The focus area of the current study is to demonstrate the robustness and feasibility of the CASL baseline boiling model for DNB performance inmore » a full 5x5 fuel bundle application. A quantitative evaluation of the DNB predictive capability is performed by comparing with corresponding experimental measurements (i.e. reference for the model validation). The reference data are provided from the Westinghouse Electricity Company (WEC). Two different grid configurations tested here include Non-Mixing Vane Grid (NMVG), and Mixing Vane Grid (MVG). Thorough validation studies with two sub-channel configurations are performed at a wide range of realistic PWR operational conditions.« less

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

On the Solidification and Structure Formation during Casting of Large Inserts in Ferritic Nodular Cast Iron

NASA Astrophysics Data System (ADS)

Tadesse, Abel; Fredriksson, Hasse

2018-06-01

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

Grid-to-rod flow-induced impact study for PWR fuel in reactor

DOE PAGES

Jiang, Hao; Qu, Jun; Lu, Roger Y.; ...

2016-06-10

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

Application of a Simple Model to Predict Environmental Radionuclide Levels and Consequential Dose Rates on the South Welsh Coast, U.K.

NASA Astrophysics Data System (ADS)

Halliwell, C. M.; McKay, W. A.

1994-02-01

The impact of liquid effluent discharges, from both existing nuclear power stations and from a possible future pressurized water reactor (PWR), on the levels of radioactivity in Welsh Severn coastal waters has been addressed in this study through the use of a simple box model. If a PWR was in operation at Hinkley Point, and assuming that the existing discharges into the estuary remained the same as in 1989, the levels of the most radiologically significant radionuclide, 137Cs, in seawater along the Welsh shoreline are predicted to increase by 7% (inner estuary), 7% (Welsh outer estuary) and 5% (inner channel) and in sediment by 0·3, 1·3 and 2% respectively. The radiation dose rate from 137Cs to members of the coastal population alone would show only a marginal increase due to these changes, and would remain less than 1% of the internationally recognized limit.

Posttest analysis of LOFT LOCE L2-3 using the ESA RELAP4 blowdown model. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perryman, J.L.; Samuels, T.K.; Cooper, C.H.

A posttest analysis of the blowdown portion of Loss-of-Coolant Experiment (LOCE) L2-3, which was conducted in the Loss-of-Fluid Test (LOFT) facility, was performed using the experiment safety analysis (ESA) RELAP4/MOD5 computer model. Measured experimental parameters were compared with the calculations in order to assess the conservatisms in the ESA RELAP4/MOD5 model.

ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walter, Matthew; Yin, Shengjun; Stevens, Gary

2012-01-01

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperaturemore » (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.« less

Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

2011-07-01

In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of amore » power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.« less

Identification of poor households for premium exemptions in Ghana's National Health Insurance Scheme: empirical analysis of three strategies.

PubMed

Aryeetey, Genevieve Cecilia; Jehu-Appiah, Caroline; Spaan, Ernst; D'Exelle, Ben; Agyepong, Irene; Baltussen, Rob

2010-12-01

To evaluate the effectiveness of three alternative strategies to identify poor households: means testing (MT), proxy means testing (PMT) and participatory wealth ranking (PWR) in urban, rural and semi-urban settings in Ghana. The primary motivation was to inform implementation of the National Health Insurance policy of premium exemptions for the poorest households. Survey of 145-147 households per setting to collect data on consumption expenditure to estimate MT measures and of household assets to estimate PMT measures. We organized focus group discussions to derive PWR measures. We compared errors of inclusion and exclusion of PMT and PWR relative to MT, the latter being considered the gold standard measure to identify poor households. Compared to MT, the errors of exclusion and inclusion of PMT ranged between 0.46-0.63 and 0.21-0.36, respectively, and of PWR between 0.03-0.73 and 0.17-0.60, respectively, depending on the setting. Proxy means testing and PWR have considerable errors of exclusion and inclusion in comparison with MT. PWR is a subjective measure of poverty and has appeal because it reflects community's perceptions on poverty. However, as its definition of the poor varies across settings, its acceptability as a uniform strategy to identify the poor in Ghana may be questionable. PMT and MT are potential strategies to identify the poor, and their relative societal attractiveness should be judged in a broader economic analysis. This study also holds relevance to other programmes that require identification of the poor in low-income countries. © 2010 Blackwell Publishing Ltd.

Trends in Global Vegetation Activity and Climatic Drivers Indicate a Decoupled Response to Climate Change

PubMed Central

Schut, Antonius G. T.; Ivits, Eva; Conijn, Jacob G.; ten Brink, Ben; Fensholt, Rasmus

2015-01-01

Detailed understanding of a possible decoupling between climatic drivers of plant productivity and the response of ecosystems vegetation is required. We compared trends in six NDVI metrics (1982–2010) derived from the GIMMS3g dataset with modelled biomass productivity and assessed uncertainty in trend estimates. Annual total biomass weight (TBW) was calculated with the LINPAC model. Trends were determined using a simple linear regression, a Thiel-Sen medium slope and a piecewise regression (PWR) with two segments. Values of NDVI metrics were related to Net Primary Production (MODIS-NPP) and TBW per biome and land-use type. The simple linear and Thiel-Sen trends did not differ much whereas PWR increased the fraction of explained variation, depending on the NDVI metric considered. A positive trend in TBW indicating more favorable climatic conditions was found for 24% of pixels on land, and for 5% a negative trend. A decoupled trend, indicating positive TBW trends and monotonic negative or segmented and negative NDVI trends, was observed for 17–36% of all productive areas depending on the NDVI metric used. For only 1–2% of all pixels in productive areas, a diverging and greening trend was found despite a strong negative trend in TBW. The choice of NDVI metric used strongly affected outcomes on regional scales and differences in the fraction of explained variation in MODIS-NPP between biomes were large, and a combination of NDVI metrics is recommended for global studies. We have found an increasing difference between trends in climatic drivers and observed NDVI for large parts of the globe. Our findings suggest that future scenarios must consider impacts of constraints on plant growth such as extremes in weather and nutrient availability to predict changes in NPP and CO2 sequestration capacity. PMID:26466347

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.

Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Pozsgai, C.

1992-01-01

Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative naturemore » of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.« less

Evaluation of a Powered Ankle-Foot Prosthesis during Slope Ascent Gait

PubMed Central

2016-01-01

Passive prosthetic feet lack active plantarflexion and push-off power resulting in gait deviations and compensations by individuals with transtibial amputation (TTA) during slope ascent. We sought to determine the effect of active ankle plantarflexion and push-off power provided by a powered prosthetic ankle-foot (PWR) on lower extremity compensations in individuals with unilateral TTA as they walked up a slope. We hypothesized that increased ankle plantarflexion and push-off power would reduce compensations commonly observed with a passive, energy-storing-returning prosthetic ankle-foot (ESR). We compared the temporal spatial, kinematic, and kinetic measures of ten individuals with TTA (age: 30.2 ± 5.3 yrs) to matched abled-bodied (AB) individuals during 5° slope ascent. The TTA group walked with an ESR and separately with a PWR. The PWR produced significantly greater prosthetic ankle plantarflexion and push-off power generation compared to an ESR and more closely matched AB values. The PWR functioned similar to a passive ESR device when transitioning onto the prosthetic limb due to limited prosthetic dorsiflexion, which resulted in similar deviations and compensations. In contrast, when transitioning off the prosthetic limb, increased ankle plantarflexion and push-off power provided by the PWR contributed to decreased intact limb knee extensor power production, lessening demand on the intact limb knee. PMID:27977681

Evaluation of a Powered Ankle-Foot Prosthesis during Slope Ascent Gait.

PubMed

Rábago, Christopher A; Aldridge Whitehead, Jennifer; Wilken, Jason M

2016-01-01

Passive prosthetic feet lack active plantarflexion and push-off power resulting in gait deviations and compensations by individuals with transtibial amputation (TTA) during slope ascent. We sought to determine the effect of active ankle plantarflexion and push-off power provided by a powered prosthetic ankle-foot (PWR) on lower extremity compensations in individuals with unilateral TTA as they walked up a slope. We hypothesized that increased ankle plantarflexion and push-off power would reduce compensations commonly observed with a passive, energy-storing-returning prosthetic ankle-foot (ESR). We compared the temporal spatial, kinematic, and kinetic measures of ten individuals with TTA (age: 30.2 ± 5.3 yrs) to matched abled-bodied (AB) individuals during 5° slope ascent. The TTA group walked with an ESR and separately with a PWR. The PWR produced significantly greater prosthetic ankle plantarflexion and push-off power generation compared to an ESR and more closely matched AB values. The PWR functioned similar to a passive ESR device when transitioning onto the prosthetic limb due to limited prosthetic dorsiflexion, which resulted in similar deviations and compensations. In contrast, when transitioning off the prosthetic limb, increased ankle plantarflexion and push-off power provided by the PWR contributed to decreased intact limb knee extensor power production, lessening demand on the intact limb knee.

Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sonat Sen; Gilles Youinou

2013-02-01

It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this casemore » the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)« less

ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

DOE Office of Scientific and Technical Information (OSTI.GOV)

2013-11-01

1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69more » rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in all respects except that it contained a partial blockage formed by attaching sleeves (or "balloons") to some of the rods. 6. SOURCE AND SCOPE OF DATA Phenomena Tested - Heat transfer in the core of a PWR during a re-flood phase of postulated large break LOCA. Test Designation - Achilles Rig. The programme includes the following types of experiments: - on an unballooned cluster: -- single phase air flow -- low pressure level swell -- low flooding rate re-flood -- high flooding rate re-flood - on a ballooned cluster containing 80% blockage formed by 16 balloon sleeves -- single phase air flow -- low flooding rate re-flood 7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM N/A 8. DATA FORMAT AND COMPUTER Many Computers (M00019MNYCP00). 9. TYPICAL RUNNING TIME N/A 11. CONTENTS OF LIBRARY The ACHILLES package contains test data and associated data processing software as well as the documentation listed above. 12. DATE OF ABSTRACT November 2013. KEYWORDS: DATABASES, BENCHMARKS, HEAT TRANSFER, LOSS-OF-COLLANT ACCIDENT, PWR REACTORS, REFLOODING« less

Performance evaluation of two-stage fuel cycle from SFR to PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fei, T.; Hoffman, E.A.; Kim, T.K.

2013-07-01

One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with anmore » average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)« less

Association between gestational weight gain according to body mass index and postpartum weight in a large cohort of Danish women.

PubMed

Rode, Line; Kjærgaard, Hanne; Ottesen, Bent; Damm, Peter; Hegaard, Hanne K

2012-02-01

Our aim was to investigate the association between gestational weight gain (GWG) and postpartum weight retention (PWR) in pre-pregnancy underweight, normal weight, overweight or obese women, with emphasis on the American Institute of Medicine (IOM) recommendations. We performed secondary analyses on data based on questionnaires from 1,898 women from the "Smoke-free Newborn Study" conducted 1996-1999 at Hvidovre Hospital, Denmark. Relationship between GWG and PWR was examined according to BMI as a continuous variable and in four groups. Association between PWR and GWG according to IOM recommendations was tested by linear regression analysis and the association between PWR ≥ 5 kg (11 lbs) and GWG by logistic regression analysis. Mean GWG and mean PWR were constant for all BMI units until 26-27 kg/m(2). After this cut-off mean GWG and mean PWR decreased with increasing BMI. Nearly 40% of normal weight, 60% of overweight and 50% of obese women gained more than recommended during pregnancy. For normal weight and overweight women with GWG above recommendations the OR of gaining ≥ 5 kg (11 lbs) 1-year postpartum was 2.8 (95% CI 2.0-4.0) and 2.8 (95% CI 1.3-6.2, respectively) compared to women with GWG within recommendations. GWG above IOM recommendations significantly increases normal weight, overweight and obese women's risk of retaining weight 1 year after delivery. Health personnel face a challenge in prenatal counseling as 40-60% of these women gain more weight than recommended for their BMI. As GWG is potentially modifiable, our study should be followed by intervention studies focusing on GW.

Recent operating experiences with steam generators in Japanese NPPs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yashima, Seiji

1997-02-01

In 1994, the Genkai-3 of Kyushu Electric Power Co., Inc. and the Ikata-3 of Shikoku Electric Power Co., Inc. started commercial operation, and now 22 PWR plants are being operated in Japan. Since the first PWR plant now 22 PWR plants are being operated in was started to operate, Japanese PWR plants have had an operating experience of approx. 280 reactor-years. During that period, many tube degradations have been experienced in steam generators (SGs). And, in 1991, the steam generator tube rupture (SGTR) occurred in the Mihama-2 of Kansai Electric Power Co., Inc. However, the occurrence of tube degradation ofmore » SGs has been decreased by the instructions of the MITI as regulatory authorities, efforts of Electric Utilities, and technical support from the SG manufacturers. Here the author describes the recent SGs in Japan about the following points. (1) Recent Operating Experiences (2) Lessons learned from Mihama-2 SGTR (3) SG replacement (4) Safety Regulations on SG (5) Research and development on SG.« less

Uniaxial low cycle fatigue behavior for pre-corroded 16MND5 bainitic steel in simulated pressurized water reactor environment

NASA Astrophysics Data System (ADS)

Chen, Xu; Ren, Bin; Yu, Dunji; Xu, Bin; Zhang, Zhe; Chen, Gang

2018-06-01

The effects of uniaxial tension properties and low cycle fatigue behavior of 16MND5 bainitic steel cylinder pre-corroded in simulated pressurized water reactor (PWR) were investigated by fatigue at room temperature in air and immersion test system, scanning electron microscopy (SEM), energy disperse spectroscopy (EDS). The experimental results indicated that the corrosion fatigue lives of 16MND5 specimen were significantly affected by the strain amplitude and simulated PWR environments. The compositions of corrosion products were complexly formed in simulated PWR environments. The porous corrosion surface of pre-corroded materials tended to generate pits as a result of promoting contact area to the fresh metal, which promoted crack initiation. For original materials, the fatigue cracks initiated at inclusions imbedded in the micro-cracks. Moreover, the simulated PWR environments degraded the mechanical properties and low cycle fatigue behavior of 16MND5 specimens remarkably. Pre-corrosion of 16MND5 specimen mainly affected the plastic term of the Coffin-Manson equation.

Design study of long-life PWR using thorium cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

2012-06-06

Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that {sup 231}Pa better than {sup 237}Np as burnable poisons in thorium fuel system. Thorium oxide system with 8%{sup 233}U enrichment and 7.6{approx} 8%{sup 231}Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1%{Delta}k/kmore » and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53%{Delta}k/k and reduced power peaking during its operation.« less

Effects of Thermo-Mechanical Treatments on Deformation Behavior and IGSCC Susceptibility of Stainless Steels in Pwr Primary Water Chemistry

NASA Astrophysics Data System (ADS)

Nouraei, S.; Tice, D. R.; Mottershead, K. J.; Wright, D. M.

Field experience of 300 series stainless steels in the primary circuit of PWR plant has been good. Stress Corrosion Cracking of components has been infrequent and mainly associated with contamination by impurities/oxygen in occluded locations. However, some instances of failures have been observed which cannot necessarily be attributed to deviations in the water chemistry. These failures appear to be associated with the presence of cold-work produced by surface finishing and/or by welding-induced shrinkage. Recent data indicate that some heats of SS show an increased susceptibility to SCC; relatively high crack growth rates were observed even when the crack growth direction is orthogonal to the cold-work direction. SCC of cold-worked SS in PWR coolant is therefore determined by a complex interaction of material composition, microstructure, prior cold-work and heat treatment. This paper will focus on the interactions between these parameters on crack propagation in simulated PWR conditions.

System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs,more » were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.« less

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculationsmore » accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)« less

Plasmon waveguide resonance sensor using an Au-MgF2 structure.

PubMed

Zhou, Yanfei; Zhang, Pengfei; He, Yonghong; Xu, Zihao; Liu, Le; Ji, Yanhong; Ma, Hui

2014-10-01

We report an Au − MgF(2) plasmon waveguide resonance (PWR) sensor in this work. The characteristics of this sensing structure are compared with a surface plasmon resonance (SPR) structure theoretically and experimentally. The transverse-magnetic-polarized PWR sensor has a refractive index resolution of 9.3 × 10(-7) RIU, which is 6 times smaller than that of SPR at the incident light wavelength of 633 nm, and the transverse-electric-polarized PWR sensor has a refractive index resolution of 3.0 × 10(-6) RIU. This high-resolution sensor is easy to build and is less sensitive to film coating deviations.

Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Calloo, A.; Vidal, J. F.; Le Tellier, R.

2012-07-01

This paper deals with the resolution of the integro-differential form of the Boltzmann transport equation for neutron transport in nuclear reactors. In multigroup theory, deterministic codes use transfer cross sections which are expanded on Legendre polynomials. This modelling leads to negative values of the transfer cross section for certain scattering angles, and hence, the multigroup scattering source term is wrongly computed. The first part compares the convergence of 'Legendre-expanded' cross sections with respect to the order used with the method of characteristics (MOC) for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section is developed using piecewise-constant functions, whichmore » better models the multigroup transfer cross section and prevents the occurrence of any negative value for it. The second part focuses on the method of solving the transport equation with the above-mentioned piecewise-constant cross sections for lattice calculations for PWR cells. This expansion thereby constitutes a 'reference' method to compare the conventional Legendre expansion to, and to determine its pertinence when applied to reactor physics calculations. (authors)« less

Bio-knowledge based filters improve residue-residue contact prediction accuracy.

PubMed

Wozniak, P P; Pelc, J; Skrzypecki, M; Vriend, G; Kotulska, M

2018-05-29

Residue-residue contact prediction through direct coupling analysis has reached impressive accuracy, but yet higher accuracy will be needed to allow for routine modelling of protein structures. One way to improve the prediction accuracy is to filter predicted contacts using knowledge about the particular protein of interest or knowledge about protein structures in general. We focus on the latter and discuss a set of filters that can be used to remove false positive contact predictions. Each filter depends on one or a few cut-off parameters for which the filter performance was investigated. Combining all filters while using default parameters resulted for a test-set of 851 protein domains in the removal of 29% of the predictions of which 92% were indeed false positives. All data and scripts are available from http://comprec-lin.iiar.pwr.edu.pl/FPfilter/. malgorzata.kotulska@pwr.edu.pl. Supplementary data are available at Bioinformatics online.

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

Annual report, FY 1979 Spent fuel and fuel pool component integrity.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion.more » A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.« less

Verification of bubble tracking method and DNS examinations of single- and two-phase turbulent channel flows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tryggvason, Gretar; Bolotnov, Igor; Fang, Jun

2017-03-30

Direct numerical simulation (DNS) has been regarded as a reliable data source for the development and validation of turbulence models along with experiments. The realization of DNS usually involves a very fine mesh that should be able to resolve all relevant turbulence scales down to Kolmogorov scale [1]. As the most computationally expensive approach compared to other CFD techniques, DNS applications used to be limited to flow studies at very low Reynolds numbers. Thanks to the tremendous growth of computing power over the past decades, the simulation capability of DNS has now started overlapping with some of the most challengingmore » engineering problems. One of those examples in nuclear engineering is the turbulent coolant flow inside reactor cores. Coupled with interface tracking methods (ITM), the simulation capability of DNS can be extended to more complicated two-phase flow regimes. Departure from nucleate boiling (DNB) is the limiting critical heat flux phenomena for the majority of accidents that are postulated to occur in pressurized water reactors (PWR) [2]. As one of the major modeling and simulation (M&S) challenges pursued by CASL, the prediction capability is being developed for the onset of DNB utilizing multiphase-CFD (M-CFD) approach. DNS (coupled with ITM) can be employed to provide closure law information for the multiphase flow modeling at CFD scale. In the presented work, research groups at NCSU and UND will focus on applying different ITM to different geometries. Higher void fraction flow analysis at reactor prototypical conditions will be performed, and novel analysis methods will be developed, implemented and verified for the challenging flow conditions.« less

TRAC-PD2 posttest analysis of the CCTF Evaluation-Model Test C1-19 (Run 38). [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Motley, F.

The results of a Transient Reactor Analysis Code posttest analysis of the Cylindral Core Test Facility Evaluation-Model Test agree very well with the results of the experiment. The good agreement obtained verifies the multidimensional analysis capability of the TRAC code. Because of the steep radial power profile, the importance of using fine noding in the core region was demonstrated (as compared with poorer results obtained from an earlier pretest prediction that used a coarsely noded model).

Application of the TEMPEST computer code for simulating hydrogen distribution in model containment structures. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trent, D.S.; Eyler, L.L.

In this study several aspects of simulating hydrogen distribution in geometric configurations relevant to reactor containment structures were investigated using the TEMPEST computer code. Of particular interest was the performance of the TEMPEST turbulence model in a density-stratified environment. Computed results illustrated that the TEMPEST numerical procedures predicted the measured phenomena with good accuracy under a variety of conditions and that the turbulence model used is a viable approach in complex turbulent flow simulation.

Steam--water mixing and system hydrodynamics program. Task 4. Quarterly progress report, October 1, 1977--December 31, 1977. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbiener, W.A.; Cudnik, R.A.; Dykhuizen, R.C.

Experimental studies were conducted in a /sup 2///sub 15/-scale model of a four-loop pressurized water reactor at pressures to 75 psia to extend the understanding of steam-water interaction phenomena and processes associated with a loss-of-coolant accident. Plenum filling studies were conducted with hydraulic communication between the cold leg and core steam supplies and hot walls, with both fixed and ramped steam flows. Comparisons of correlational fits have been made for penetration data obtained with hydraulic communication, fixed cold leg steam, and no cold leg steam. Statistical tests applied to these correlational fits have indicated that the hydraulic communication and fixedmore » cold leg steam data can be considered to be a common data set. Comparing either of these data sets to the no cold leg steam data using the statistical test indicated that it was unlikely that these sets could be considered to be a common data set. The introduction of cold leg steam results in a slight decrease in penetration relative to that obtained without cold leg steam at the same value of subcooling of water entering the downcomer. A dimensionless parameter which is a weighted mean of a modified Froude number and the Weber number has been proposed as a scaling parameter for penetration data. This parameter contains an additional degree of freedom which allows data from different scales to collapse more closely to a single curve than current scaling parameters permit.« less

Optimization of small long-life PWR based on thorium fuel

NASA Astrophysics Data System (ADS)

Subkhi, Moh Nurul; Suud, Zaki; Waris, Abdul; Permana, Sidik

2015-09-01

A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity.

78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-13

... (iPWR). This guidance applies to environmental reviews associated with iPWR applications for limited... received on or before this date. ADDRESSES: You may submit comments by any of the following methods (unless... this document. You may access publicly-available information related to this document by any of the...

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code thatmore » is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason Hales; Various

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale codemore » that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Neutron Collar Evolution and Fresh PWR Assembly Measurements with a New Fast Neutron Passive Collar

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard Olsen; Geist, William H.; Root, Margaret A.

The passive neutron collar approach removes the effect of poison rods when using a 1mm Gd liner. This project sets out to solve the following challenges: BWR fuel assemblies have less mass and less neutron multiplication than PWR; and effective removal of cosmic ray spallation neutron bursts needed via QC tests.

MC21 analysis of the MIT PWR benchmark: Hot zero power results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly Iii, D. J.; Aviles, B. N.; Herman, B. R.

2013-07-01

MC21 Monte Carlo results have been compared with hot zero power measurements from an operating pressurized water reactor (PWR), as specified in a new full core PWR performance benchmark from the MIT Computational Reactor Physics Group. Included in the comparisons are axially integrated full core detector measurements, axial detector profiles, control rod bank worths, and temperature coefficients. Power depressions from grid spacers are seen clearly in the MC21 results. Application of Coarse Mesh Finite Difference (CMFD) acceleration within MC21 has been accomplished, resulting in a significant reduction of inactive batches necessary to converge the fission source. CMFD acceleration has alsomore » been shown to work seamlessly with the Uniform Fission Site (UFS) variance reduction method. (authors)« less

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaver, Mark W.; Lanning, Donald D.

2010-02-01

The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum ofmore » the individual components equaling the measured values.« less

Optimization of small long-life PWR based on thorium fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subkhi, Moh Nurul, E-mail: nsubkhi@students.itb.ac.id; Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung Djati Bandung Jalan A.H Nasution 105 Bandung; Suud, Zaki, E-mail: szaki@fi.itb.ac.id

2015-09-30

A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {supmore » 231}Pa give low excess reactivity.« less

CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferrer, R.; Rhodes, J.; Smith, K.

2012-07-01

The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, P.; Umminger, K.J.; Schoen, B.

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where themore » decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).« less

Development a computer codes to couple PWR-GALE output and PC-CREAM input

NASA Astrophysics Data System (ADS)

Kuntjoro, S.; Budi Setiawan, M.; Nursinta Adi, W.; Deswandri; Sunaryo, G. R.

2018-02-01

Radionuclide dispersion analysis is part of an important reactor safety analysis. From the analysis it can be obtained the amount of doses received by radiation workers and communities around nuclear reactor. The radionuclide dispersion analysis under normal operating conditions is carried out using the PC-CREAM code, and it requires input data such as source term and population distribution. Input data is derived from the output of another program that is PWR-GALE and written Population Distribution data in certain format. Compiling inputs for PC-CREAM programs manually requires high accuracy, as it involves large amounts of data in certain formats and often errors in compiling inputs manually. To minimize errors in input generation, than it is make coupling program for PWR-GALE and PC-CREAM programs and a program for writing population distribution according to the PC-CREAM input format. This work was conducted to create the coupling programming between PWR-GALE output and PC-CREAM input and programming to written population data in the required formats. Programming is done by using Python programming language which has advantages of multiplatform, object-oriented and interactive. The result of this work is software for coupling data of source term and written population distribution data. So that input to PC-CREAM program can be done easily and avoid formatting errors. Programming sourceterm coupling program PWR-GALE and PC-CREAM is completed, so that the creation of PC-CREAM inputs in souceterm and distribution data can be done easily and according to the desired format.

Shuttle Engine Designs Revolutionize Solar Power

NASA Technical Reports Server (NTRS)

2014-01-01

The Space Shuttle Main Engine was built under contract to Marshall Space Flight Center by Rocketdyne, now part of Pratt & Whitney Rocketdyne (PWR). PWR applied its NASA experience to solar power technology and licensed the technology to Santa Monica, California-based SolarReserve. The company now develops concentrating solar power projects, including a plant in Nevada that has created 4,300 jobs during construction.

Fourier Transform-Plasmon Waveguide Spectroscopy: A Nondestructive Multifrequency Method for Simultaneously Determining Polymer Thickness and Apparent Index of Refraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bobbitt, Jonathan M; Weibel, Stephen C; Elshobaki, Moneim

2014-12-16

Fourier transform (FT)-plasmon waveguide resonance (PWR) spectroscopy measures light reflectivity at a waveguide interface as the incident frequency and angle are scanned. Under conditions of total internal reflection, the reflected light intensity is attenuated when the incident frequency and angle satisfy conditions for exciting surface plasmon modes in the metal as well as guided modes within the waveguide. Expanding upon the concept of two-frequency surface plasmon resonance developed by Peterlinz and Georgiadis [ Opt. Commun. 1996, 130, 260], the apparent index of refraction and the thickness of a waveguide can be measured precisely and simultaneously by FT-PWR with an averagemore » percent relative error of 0.4%. Measuring reflectivity for a range of frequencies extends the analysis to a wide variety of sample compositions and thicknesses since frequencies with the maximum attenuation can be selected to optimize the analysis. Additionally, the ability to measure reflectivity curves with both p- and s-polarized light provides anisotropic indices of refraction. FT-PWR is demonstrated using polystyrene waveguides of varying thickness, and the validity of FT-PWR measurements are verified by comparing the results to data from profilometry and atomic force microscopy (AFM).« less

Fourier transform-plasmon waveguide spectroscopy: a nondestructive multifrequency method for simultaneously determining polymer thickness and apparent index of refraction.

PubMed

Bobbitt, Jonathan M; Weibel, Stephen C; Elshobaki, Moneim; Chaudhary, Sumit; Smith, Emily A

2014-12-16

Fourier transform (FT)-plasmon waveguide resonance (PWR) spectroscopy measures light reflectivity at a waveguide interface as the incident frequency and angle are scanned. Under conditions of total internal reflection, the reflected light intensity is attenuated when the incident frequency and angle satisfy conditions for exciting surface plasmon modes in the metal as well as guided modes within the waveguide. Expanding upon the concept of two-frequency surface plasmon resonance developed by Peterlinz and Georgiadis [Opt. Commun. 1996, 130, 260], the apparent index of refraction and the thickness of a waveguide can be measured precisely and simultaneously by FT-PWR with an average percent relative error of 0.4%. Measuring reflectivity for a range of frequencies extends the analysis to a wide variety of sample compositions and thicknesses since frequencies with the maximum attenuation can be selected to optimize the analysis. Additionally, the ability to measure reflectivity curves with both p- and s-polarized light provides anisotropic indices of refraction. FT-PWR is demonstrated using polystyrene waveguides of varying thickness, and the validity of FT-PWR measurements are verified by comparing the results to data from profilometry and atomic force microscopy (AFM).

Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

NASA Astrophysics Data System (ADS)

Hartini, Entin; Andiwijayakusuma, Dinan

2014-09-01

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id

2014-09-30

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuelmore » type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.« less

Test prediction for the German PKL Test K5A using RELAP4/MOD6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Y.S.; Haigh, W.S.; Sullivan, L.H.

RELAP4/MOD6 is the most recent modification in the series of RELAP4 computer programs developed to describe the thermal-hydraulic conditions attendant to postulated transients in light water reactor systems. The major new features in RELAP4/MOD6 include best-estimate pressurized water reactor (PWR) reflood transient analytical models for core heat transfer, local entrainment, and core vapor superheat, and a new set of heat transfer correlations for PWR blowdown and reflood. These new features were used for a test prediction of the Kraftwerk Union three-loop PRIMAR KREISLAUF (PKL) Reflood Test K5A. The results of the prediction were in good agreement with the experimental thermalmore » and hydraulic system data. Comparisons include heater rod surface temperature, system pressure, mass flow rates, and core mixture level. It is concluded that RELAP4/MOD6 is capable of accurately predicting transient reflood phenomena in the 200% cold-leg break test configuration of the PKL reflood facility.« less

Improved Biomolecular Thin-Film Sensor based on Plasmon Waveguide Resonance

NASA Astrophysics Data System (ADS)

Byard, Courtney; Aslan, Mustafa; Mendes, Sergio

2009-05-01

The design, fabrication, and characterization of a plasmon waveguide resonance (PWR) sensor are presented. Glass substrates are coated with a 35 nm gold film using electron beam evaporation, and then covered with a 143 nm aluminum oxide waveguide using an atomic layer deposition process, creating a smooth, highly transparent dielectric film. When probed in the Kretschmann configuration, the structure allows for an efficient conversion of an incident optical beam into a surface wave, which is mainly confined in the dielectric layer and exhibits a deep and narrow angular resonance. The performance (reflectance vs. incidence angle in TE polarization) is modeled using a transfer-matrix approach implemented into a Mathematica code. Our simulations and experimental data are compared with that of surface plasmon resonance (SPR) sensor using the same criteria. We show that the resolution of PWR is approximately ten times better than SPR, opening opportunities for more sensitive studies in various applications including research in protein interactions, pharmaceutical drug development, and food analysis.

Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taniguchi, T.; Ligon, D.; Stamatelatos, M.

1983-04-01

Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights inmore » the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.« less

Severe accident modeling of a PWR core with different cladding materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, S. C.; Henry, R. E.; Paik, C. Y.

2012-07-01

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCSmore » rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)« less

Cyclic and SCC Behavior of Alloy 690 HAZ in a PWR Environment

NASA Astrophysics Data System (ADS)

Alexandreanu, Bogdan; Chen, Yiren; Natesan, Ken; Shack, Bill

The objective of this work is to determine the cyclic and stress corrosion cracking (SCC) crack growth rates (CGRs) in a simulated PWR water environment for Alloy 690 heat affected zone (HAZ). In order to meet the objective, an Alloy 152 J-weld was produced on a piece of Alloy 690 tubing, and the test specimens were aligned with the HAZ. The environmental enhancement of cyclic CGRs for Alloy 690 HAZ was comparable to that measured for the same alloy in the as-received condition. The two Alloy 690 HAZ samples tested exhibited maximum SCC CGR rates of 10-11 m/s in the simulated PWR environment at 320°C, however, on average, these rates are similar or only slightly higher than those for the as-received alloy.

Analysis of crack initiation and growth in the high level vibration test at Tadotsu

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kassir, M.K.; Park, Y.J.; Hofmayer, C.H.

1993-08-01

The High Level Vibration Test data are used to assess the accuracy and usefulness of current engineering methodologies for predicting crack initiation and growth in a cast stainless steel pipe elbow under complex, large amplitude loading. The data were obtained by testing at room temperature a large scale modified model of one loop of a PWR primary coolant system at the Tadotsu Engineering Laboratory in Japan. Fatigue crack initiation time is reasonably predicted by applying a modified local strain approach (Coffin-Mason-Goodman equation) in conjunction with Miner`s rule of cumulative damage. Three fracture mechanics methodologies are applied to investigate the crackmore » growth behavior observed in the hot leg of the model. These are: the {Delta}K methodology (Paris law), {Delta}J concepts and a recently developed limit load stress-range criterion. The report includes a discussion on the pros and cons of the analysis involved in each of the methods, the role played by the key parameters influencing the formulation and a comparison of the results with the actual crack growth behavior observed in the vibration test program. Some conclusions and recommendations for improvement of the methodologies are also provided.« less

Methods and benefits of experimental seismic evaluation of nuclear power plants. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-07-01

This study reviews experimental techniques, instrumentation requirements, safety considerations, and benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models. Techniques for identification of resonant frequencies, damping, and mode shapes, are discussed. The benefits of testing with regard to increased damping and more accurate computer models are oulined. A test plan, schedule and budget are presented for a typical PWR nuclear power plant.

Reflux cooling experiments on the NCSU scaled PWR facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doster, J.M.; Giavedoni, E.

1993-01-01

Under loss of forced circulation, coupled with the loss or reduction in primary side coolant inventory, horizontal stratified flows can develop in the hot and cold legs of pressurized water reactors (PWRs). Vapor produced in the reactor vessel is transported through the hot leg to the steam generator tubes where it condenses and flows back to the reactor vessel. Within the steam generator tubes, the flow regimes may range from countercurrent annular flow to single-phase convection. As a result, a number of heat transfer mechanisms are possible, depending on the loop configuration, total heat transfer rate, and the steam flowmore » rate within the tubes. These include (but are not limited to) two-phase natural circulation, where the condensate flows concurrent to the vapor stream and is transported to the cold leg so that the entire reactor coolant loop is active, and reflux cooling, where the condensate flows back down the interior of the coolant tubes countercurrent to the vapor stream and is returned to the reactor vessel through the hot leg. While operating in the reflux cooling mode, the cold leg can effectively be inactive. Heat transfer can be further influenced by noncondensables in the vapor stream, which accumulate within the upper regions of the steam generator tube bundle. In addition to reducing the steam generator's effective heat transfer area, under these conditions operation under natural circulation may not be possible, and reflux cooling may be the only viable heat transfer mechanism. The scaled PWR (SPWR) facility in the nuclear engineering department at North Carolina State Univ. (NCSU) is being used to study the effectiveness of two-phase natural circulation and reflux cooling under conditions associated with loss of forced circulation, midloop coolant levels, and noncondensables in the primary coolant system.« less

Postpartum weight trajectories in overweight and lean women.

PubMed

Bogaerts, Annick; De Baetselier, Elyne; Ameye, Lieveke; Dilles, Tinne; Van Rompaey, Bart; Devlieger, Roland

2017-06-01

overweight and obesity in women of reproductive age are increasing and are often linked with excessive weight gain in pregnancy and weight retention after birth. Studies on spontaneous maternal weight trajectory after childbirth are scarce. we describe women's spontaneous weight trajectory during the first six weeks of the postpartum period and its relationship between Body Mass Index and socio-demographical, behavioural and psychological variables. data from 212 women who gave birth in three regional hospitals were collected prospectively between December 2015 and February 2016. Potential determinants were examined during pregnancy and the postpartum period at four and six weeks after childbirth. Descriptive statistics and a linear multivariate regression model were used. Early postnatal weight retention (PWR) was defined as the difference between the maternal weight six weeks after childbirth and the pre-pregnancy weight (kg). mean PWR at six weeks after childbirth was 3.3kg (SD 4.1), with a range between -7 and +16.2kg; 81% reported some weight retention (PWR>0kg), and 36% showed a high weight retention (PWR≥5kg). Women with a BMI <25kg/m 2 showed a significantly higher mean PWR six weeks after childbirth compared to women with a BMI ≥25kg/m 2 (4.0kg versus 1.6kg, p=0.002). There was a significant correlation between maternal weight retention and gestational weight gain (GWG) (B=0.65, p<0.001) and pre-pregnancy body mass index <25kg/m 2 (B=1.12, p=0.017), six weeks after childbirth. weight retention six weeks after childbirth is associated with pre-pregnancy BMI and GWG, but contrary to expectations, lean women with excessive GWG tended to retain most weight after childbirth. No significant associations with several socio-demographical, behavioural and psychological variables were found. weight management strategies around pregnancy should not be limited to overweight and obese mothers. Women with pre-pregnancy BMI <25kg/m 2 require equal attention to prevent postnatal weight retention. Copyright © 2016 Elsevier Ltd. All rights reserved.

Physics of hydride fueled PWR

NASA Astrophysics Data System (ADS)

Ganda, Francesco

The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barry, Kenneth

The Nuclear Energy Institute (NEI) Small Modular Reactor (SMR) Licensing Task Force (TF) has been evaluating licensing issues unique and important to iPWRs, ranking these issues, and developing NEI position papers for submittal to the U.S. Nuclear Regulatory Commission (NRC) during the past three years. Papers have been developed and submitted to the NRC in a range of areas including: Price-Anderson Act, NRC annual fees, security, modularity, and staffing. In December, 2012, NEI completed a draft position paper on SMR source terms and participated in an NRC public meeting presenting a summary of this paper, which was subsequently submitted tomore » the NRC. One important conclusion of the source term paper was the evaluation and selection of high importance areas where additional research would have a significant impact on source terms. The highest ranked research area was iPWR containment aerosol natural deposition. The NRC accepts the use of existing aerosol deposition correlations in Regulatory Guide 1.183, but these were developed for large light water reactor (LWR) containments. Application of these correlations to an iPWR design has resulted in greater than a ten-fold reduction of containment airborne aerosol inventory as compared to large LWRs. Development and experimental justification of containment aerosol natural deposition correlations specifically for the unique iPWR containments is expected to result in a large reduction of design basis and beyond-design-basis accident source terms with concomitantly smaller dose to workers and the public. Therefore, NRC acceptance of iPWR containment aerosol natural deposition correlations will directly support the industry’s goal of reducing the Emergency Planning Zone (EPZ) for SMRs. Based on the results in this work, it is clear that thermophoresis is relatively unimportant for iPWRs. Gravitational settling is well understood, and may be the dominant process for a dry environment. Diffusiophoresis and enhanced settling by particle growth are the dominant processes for determining DFs for expected conditions in an iPWR containment. These processes are dependent on the areato-volume (A/V) ratio, which should benefit iPWR designs because these reactors have higher A/Vs compared to existing LWRs.« less

New core-reflector boundary conditions for transient nodal reactor calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, E.K.; Kim, C.H.; Joo, H.K.

1995-09-01

New core-reflector boundary conditions designed for the exclusion of the reflector region in transient nodal reactor calculations are formulated. Spatially flat frequency approximations for the temporal neutron behavior and two types of transverse leakage approximations in the reflector region are introduced to solve the transverse-integrated time-dependent one-dimensional diffusion equation and then to obtain relationships between net current and flux at the core-reflector interfaces. To examine the effectiveness of new core-reflector boundary conditions in transient nodal reactor computations, nodal expansion method (NEM) computations with and without explicit representation of the reflector are performed for Laboratorium fuer Reaktorregelung und Anlagen (LRA) boilingmore » water reactor (BWR) and Nuclear Energy Agency Committee on Reactor Physics (NEACRP) pressurized water reactor (PWR) rod ejection kinetics benchmark problems. Good agreement between two NEM computations is demonstrated in all the important transient parameters of two benchmark problems. A significant amount of CPU time saving is also demonstrated with the boundary condition model with transverse leakage (BCMTL) approximations in the reflector region. In the three-dimensional LRA BWR, the BCMTL and the explicit reflector model computations differ by {approximately}4% in transient peak power density while the BCMTL results in >40% of CPU time saving by excluding both the axial and the radial reflector regions from explicit computational nodes. In the NEACRP PWR problem, which includes six different transient cases, the largest difference is 24.4% in the transient maximum power in the one-node-per-assembly B1 transient results. This difference in the transient maximum power of the B1 case is shown to reduce to 11.7% in the four-node-per-assembly computations. As for the computing time, BCMTL is shown to reduce the CPU time >20% in all six transient cases of the NEACRP PWR.« less

Spelling Well Despite Developmental Language Disorder: What Makes it Possible?

PubMed Central

Rakhlin, Natalia; Cardoso-Martins, Cláudia; Kornilov, Sergey A.; Grigorenko, Elena L.

2013-01-01

The goal of the study was to investigate the overlap between Developmental Language Disorder (DLD) and Developmental Dyslexia, identified through spelling difficulties (SD), in Russian-speaking children. In particular, we studied the role of phoneme awareness (PA), rapid automatized naming (RAN), pseudoword repetition (PWR), morphological (MA) and orthographic awareness (OA) in differentiating between children with DLD who have SD from children with DLD who are average spellers by comparing the two groups to each other, to typically developing children as well as children with SD but without spoken language deficits. One hundred forty nine children, aged 10.40 to 14.00, participated in the study. The results indicated that the SD, DLD, and DLD/SD groups did not differ from each other on PA and RAN Letters and underperformed in comparison to the control groups. However, whereas the children with written language deficits (SD and DLD/SD groups) underperformed on RAN Objects and Digits, PWR, OA and MA, the children with DLD and no SD performed similarly to the children from the control groups on these measures. In contrast, the two groups with spoken language deficits (DLD and DLD/SD) underperformed on RAN Colors in comparison to the control groups and the group of children with SD only. The results support the notion that those children with DLD who have unimpaired PWR and RAN skills are able to overcome their weaknesses in spoken language and PA and acquire basic literacy on a par with their age peers with typical language. We also argue that our findings support a multifactorial model of developmental language disorders (DLD). PMID:23860907

Report on the PWR-radiation protection/ALARA Committee

DOE Office of Scientific and Technical Information (OSTI.GOV)

Malone, D.J.

1995-03-01

In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and informationmore » relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.« less

Evaluation of on-line chelant addition to PWR steam generators. Steam generator cleaning project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tvedt, T.J.; Wallace, S.L.; Griffin, F. Jr.

1983-09-01

The investigation of chelating agents for continuous water treatment of secondary loops of PWR steam generators were conducted in two general areas: the study of the chemistry of chelating agents and the study of materials compatability with chelating agents. The thermostability of both EDTA and HEDTA metal chelates in All Volatile Treatment (AVT) water chemistry were shown to be greater than or equal to the thermostability of EDTA metal chelates in phosphate-sulfite water chemistry. HEDTA metal chelates were shown to have a much greater stability than EDTA metal chelates. Using samples taken from the EDTA metal chelate thermostability study andmore » from the Commonwealth Research Corporation (CRC) model steam generators (MSG), EDTA decomposition products were determined. Active metal surfaces were shown to become passivated when exposed to EDTA and HEDTA concentrations as high as 0.1% w/w in AVT. Trace amounts of iron in the water were found to increase the rate of passivation. Material balance and visual inspection data from CRC model steam generators showed that metal was transported through and cleaned from the MSG's. The Inconel 600 tubes of the salt water fouled model steam generators experienced pitting corrosion. Results of this study demonstrates the feasibility of EDTA as an on-line water treatment additive to maintain nuclear steam generators in a clean condition.« less

Primary water chemistry improvement for radiation exposure reduction at Japanese PWR Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nishizawa, Eiichi

1995-03-01

Radiation exposure during the refueling outages at Japanese Pressurized Water Reactor (PWR) Plants has been gradually decreased through continuous efforts keeping the radiation dose rates at relatively low level. The improvement of primary water chemistry in respect to reduction of the radiation sources appears as one of the most important contributions to the achieved results and can be classified by the plant operation conditions as follows

Comparison of Measures of Vibration Affecting Occupants of Military Vehicles

DTIC Science & Technology

1986-12-01

8217 ,, l I WES equipment 27. The WES equipment consisted of a battery operated absorbed power ( ABS -PW) meter with signal conditioning...West Germany. These will be referred to as the ISO ride meter and the ABS -PWR ridemeter, respectively. The first implemented the vibration measure...the ABS -PWR algorithms were used with each acceleration signal source (analog and digital) to provide a comprehensive basis for comparing the vibration

Fatigue crack growth rates in a pressure vessel steel under various conditions of loading and the environment

NASA Astrophysics Data System (ADS)

Hicks, P. D.; Robinson, F. P. A.

1986-10-01

Corrosion fatigue (CF) tests have been carried out on SA508 Cl 3 pressure vessel steel, in simulated P.W.R. environments. The test variables investigated included air and P.W.R. water environments, frequency variation over the range 1 Hz to 10 Hz, transverse and longitudinal crack growth directions, temperatures of 20 °C and 50 °C, and R-ratios of 0.2 and 0.7. It was found that decreasing the test frequency increased fatigue crack growth rates (FCGR) in P.W.R. environments, P.W.R. environment testing gave enhanced crack growth (vs air tests), FCGRs were greater for cracks growing in the longitudinal direction, slight increases in temperature gave noticeable accelerations in FCGR, and several air tests gave FCGR greater than those predicted by the existing ASME codes. Fractographic evidence indicates that FCGRs were accelerated by a hydrogen embrittlement mechanism. The presence of elongated MnS inclusions aided both mechanical fatigue and hydrogen embrittlement processes, thus producing synergistically fast FCGRs. Both anodic dissolution and hydrogen embrittlement mechanisms have been proposed for the environmental enhancement of crack growth rates. Electrochemical potential measurements and potentiostatic tests have shown that sample isolation of the test specimens from the clevises in the apparatus is not essential during low temperature corrosion fatigue testing.

Statistical evaluation of the metallurgical test data in the ORR-PSF-PVS irradiation experiment. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stallmann, F.W.

1984-08-01

A statistical analysis of Charpy test results of the two-year Pressure Vessel Simulation metallurgical irradiation experiment was performed. Determination of transition temperature and upper shelf energy derived from computer fits compare well with eyeball fits. Uncertainties for all results can be obtained with computer fits. The results were compared with predictions in Regulatory Guide 1.99 and other irradiation damage models.

SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

J.W. Davis

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

Astronaut Robinson presents 2010 Silver Snoopy awards

NASA Image and Video Library

2010-06-23

NASA's John C. Stennis Space Center Director Patrick Scheuermann and astronaut Steve Robinson stand with recipients of the 2010 Silver Snoopy awards following a June 23 ceremony. Sixteen Stennis employees received the astronauts' personal award, which is presented by a member of the astronaut corps representing its core principles for outstanding flight safety and mission success. This year's recipients and ceremony participants were: (front row, l to r): Cliff Arnold (NASA), Wendy Holladay (NASA), Kendra Moran (Pratt & Whitney Rocketdyne), Mary Johnson (Jacobs Technology Facility Operating Services Contract group), Cory Beckemeyer (PWR), Dean Bourlet (PWR), Cecile Saltzman (NASA), Marla Carpenter (Jacobs FOSC), David Alston (Jacobs FOSC); (back row, l to r) Scheuermann, Don Wilson (A2 Research), Tim White (NASA), Ira Lossett (Jacobs Technology NASA Test Operations Group), Kerry Gallagher (Jacobs NTOG); Rene LeFrere (PWR), Todd Ladner (ASRC Research and Technology Solutions) and Thomas Jacks (NASA).

TRAC posttest calculations of Semiscale Test S-06-3. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ireland, J.R.; Bleiweis, P.B.

A comparison of Transient Reactor Analysis Code (TRAC) steady-state and transient results with Semiscale Test S-06-3 (US Standard Problem 8) experimental data is discussed. The TRAC model used employs fewer mesh cells than normal data comparison models so that TRAC's ability to obtain reasonable results with less computer time can be assessed. In general, the TRAC results are in good agreement with the data and the major phenomena found in the experiment are reproduced by the code with a substantial reduction in computing times.

Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Visosky, M.; Hejzlar, P.; Kazimi, M.

2006-07-01

CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining themore » TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)« less

Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parisi, Carlo; Prescott, Steve; Ma, Zhegang

This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA modelsmore » for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.« less

A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, T.Y.; Bentz, J.; Simpson, R.

1997-02-01

The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF programmore » investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.« less

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-01

This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

McConnell, Paul E.; Koenig, Greg John; Uncapher, William Leonard

2016-05-12

This report describes the third set of tests (the “DCL a shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Effects of crack tip plastic zone on corrosion fatigue cracking of alloy 690(TT) in pressurized water reactor environments

NASA Astrophysics Data System (ADS)

Xiao, J.; Qiu, S. Y.; Chen, Y.; Fu, Z. H.; Lin, Z. X.; Xu, Q.

2015-01-01

Alloy 690(TT) is widely used for steam generator tubes in pressurized water reactor (PWR), where it is susceptible to corrosion fatigue. In this study, the corrosion fatigue behavior of Alloy 690(TT) in simulated PWR environments was investigated. The microstructure of the plastic zone near the crack tip was investigated and labyrinth structures were observed. The relationship between the crack tip plastic zone and fatigue crack growth rates and the environment factor Fen was illuminated.

Waterside corrosion of Zircaloy-clad fuel rods in a PWR environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garzarolli, F.; Jorde, D.; Manzel, R.

A data base of Zircaloy corrosion behavior under PWR operating conditions has been established from previously published reports as well as from new Kraftwerk Union (KWU) fuel examinations. The data show that the reactor environment increases the corrosion. ZrO/sub 2/ film thermal conductivity is another major factor that influences corrosion behavior. It was inferred from KWU film thickness data that the oxide film thermal conductivity may decrease once circumferential cracks develop in the layer. 57 refs.

Chemical Agonists of the PML/Daxx Pathway for Prostate Cancer Therapy

DTIC Science & Technology

2011-04-01

positive nuclei. These data suggest that the assay is highly specific and will not suffer from promiscuous reactivity with NIH library compounds...Figure 16B). Strikingly, when we compared Daxx levels in PCa cell lines to a nontumorigenic human prostatic epithelial line, PWR -1E, they were...Lysates from six different cell types ( PWR -1E, ALVA-31 Daxx K/D, ALVA-31 WT, DU145, LNCaP, and PC3) were normalized for total protein content (60 μg

Vectorized and multitasked solution of the few-group neutron diffusion equations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zee, S.K.; Turinsky, P.J.; Shayer, Z.

1989-03-01

A numerical algorithm with parallelism was used to solve the two-group, multidimensional neutron diffusion equations on computers characterized by shared memory, vector pipeline, and multi-CPU architecture features. Specifically, solutions were obtained on the Cray X/MP-48, the IBM-3090 with vector facilities, and the FPS-164. The material-centered mesh finite difference method approximation and outer-inner iteration method were employed. Parallelism was introduced in the inner iterations using the cyclic line successive overrelaxation iterative method and solving in parallel across lines. The outer iterations were completed using the Chebyshev semi-iterative method that allows parallelism to be introduced in both space and energy groups. Formore » the three-dimensional model, power, soluble boron, and transient fission product feedbacks were included. Concentrating on the pressurized water reactor (PWR), the thermal-hydraulic calculation of moderator density assumed single-phase flow and a closed flow channel, allowing parallelism to be introduced in the solution across the radial plane. Using a pinwise detail, quarter-core model of a typical PWR in cycle 1, for the two-dimensional model without feedback the measured million floating point operations per second (MFLOPS)/vector speedups were 83/11.7. 18/2.2, and 2.4/5.6 on the Cray, IBM, and FPS without multitasking, respectively. Lower performance was observed with a coarser mesh, i.e., shorter vector length, due to vector pipeline start-up. For an 18 x 18 x 30 (x-y-z) three-dimensional model with feedback of the same core, MFLOPS/vector speedups of --61/6.7 and an execution time of 0.8 CPU seconds on the Cray without multitasking were measured. Finally, using two CPUs and the vector pipelines of the Cray, a multitasking efficiency of 81% was noted for the three-dimensional model.« less

High-temperature Gas Reactor (HTGR)

NASA Astrophysics Data System (ADS)

Abedi, Sajad

2011-05-01

General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

Review of PWR fuel rod waterside corrosion behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garzarolli, F.; Jorde, D.; Manzel, R.

Waterside corrosion of Zircaloy has generally not been a problem under normal PWR operating conditions, although some instances of accelerated corrosion have been reported. However, an incentive exists to extend the average fuel rod discharge burnups to about 50,000 MWd/MTU. To minimize corrosion at these extended burnups, the factors which influence Zircaloy corrosion need to be better understood. A data base of Zircaloy corrosion behavior under PWR operating conditions has been established. The data are compiled previously published reports as well as from new Kraftwerk Union examinations. A non-destructive eddy-current technique is used to measure the oxide layer thickness onmore » fuel rods. Comparisons of measuremnts made using this eddy-current technique with those made by usual metallographic methods indicate good agreement. The data were evaluated by defining a fitting factor F which describes the increase in corrosion rate observed in-reactor over that observed from measurements of ex-reactor corrosion coupons.« less

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

DOE PAGES

George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; ...

2014-09-29

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Furthermore, implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding.« less

Derivation of the Korean radwaste scaling factor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwang Yong Jee; Hong Joo Ahn; Se Chul Sohn

2007-07-01

The concentrations of several radionuclides in low and intermediate level radioactive waste (LILW) drums have to be determined before shipping to disposal facilities. A notice, by the Ministry of Science and Technology (MOST) of the Korean Government, related to the disposal of LILW drums came into effect at the beginning of 2005, with regards to a radionuclide regulation inside a waste drum. MOST allows for an indirect radionuclide assay using a scaling factor to measure the inventories due to the difficulty of nondestructively measuring the essential {alpha} and {beta}-emitting nuclides inside a drum. That is, a scaling factor calculated throughmore » a correlation of the {alpha} or {beta}-emitting nuclide (DTM, Difficult-To-Measure) with a {gamma}-emitting nuclide (ETM, Easy-To-Measure) which has systematically similar properties with DTM nuclides. In this study, radioactive wastes, such as spent resin and dry active waste which were generated at different sites of a PWR and a site of a PHWR type Korean NPP, were partially sampled and analyzed for regulated radionuclides by using radiochemical methods. According to a reactor type and a waste form, the analysis results of each radionuclide were classified. Korean radwaste scaling factor was derived from database of radionuclide concentrations. (authors)« less

Developing a United States Marine Corps Organizational and Intermediate Level Maintenance Performance Cost Model

DTIC Science & Technology

2009-12-01

analysts. Material readiness is an issue that affects the entire Marine Corps, so the Marine Corps transitioned from a reactive to a proactive life cycle...model. Table 11 also includes ̂ (rho- hat), which takes into account the serial correlation of the residual at time t and time t-1. 57 Parameter ...TRL Mounted AS-4429D AS-4429D A0153 9 -- Radio Set, AN/MRC-142C A0170 19 -- AN/TSQ 226 (V)3 Trojan Spirit AN/TSQ 226 A0172 7 HZ DDS-R/M PWR Module (PM

Pretest analysis of Semiscale Mod-3 baseline test S-07-8 and S-07-9

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fineman, C.P.; Steiner, J.L.; Snider, D.M.

This document contains a pretest analysis of the Semiscale Mod-3 system thermal-hydraulic response for the second and third integral tests in Test Series 7 (Tests S-07-8 and S-07-9). Test Series 7 is the first test series to be conducted with the Semiscale Mod-3 system. The design of the Mod-3 system includes an improved representation of certain portions of a pressurized water reactor (PWR) when compared to the previously operated Semiscale Mod-1 system. The improvements include a new vessel which contains a full length (3.66 m) core, a full length upper plenum and upper head, and an external downcomer. An activemore » pump and active steam generator scaled to their pressurized water reactor (PWR) counterparts have been added to the broken loop. The upper head design includes the capability to simulate emergency core coolant (ECC) injection into this region. Test Series 7 is divided into three groups of tests that emphasize the evaluation of the Mod-3 system performance during different phases of the loss-of-coolant experiment (LOCE) transient. The last test group, which includes Tests S-07-8 and S-07-9, will be used to evaluate the integral behavior of the system. The previous two test groups were used to evaluate the blowdown behavior and the reflood behavior of the system. 3 refs., 35 figs., 12 tabs.« less

Cooling of core debris and the impact on containment pressure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, J.W.

1981-07-01

An evaluation of the core debris/water interactions associated with a postulated meltdown of a PWR and its impact on the containment pressure is presented. In the event of a complete core meltdown in a PWR, the interaction of molten debris with water in the bottom head of the reactor vessel could result in complete evaporation of water and breach of the vessel wall. In the reactor cavity, the debris-water interaction may lead to a rapid generation of steam, which could lead to pressures beyond the containment building limit. Previous analysis of the debris-water interactions with the MARCH code was basedmore » on the single-sphere model, in which the internal and surface heat transfer are the controlling mechanisms. In this study, the potential in-vessel and ex-vessel debris-water interactions are analyzed in terms of porous debris bed models. The debris cooling and steam generation are controlled by the hydrodynamics of the two-phase flow. The porous models developed by Dhir-Catton and by Lipinski were examined and used to test their impact on containment dynamics. The tests include several particle sizes from 1 mm to 50 mm. Detailed transient data on the pressure, temperature, and mass of steam in the containment building was obtained for all cases. Bands of pressure variation which represents the possible pressure rise under accident conditions were obtained for the Dhir-Catton model and for the Lipinski model. The results show that, for the case of a wet cavity, the magnitude of the predicted pressure rises is not strongly affected by the different models. The occurrence of the peak pressure, however, is considerably delayed by using the debris bed model. For the case of a dry cavity, a large reduction of the peak pressure is obtained by using the debris bed model.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Faidy, C.

Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

Planar Monolithic Schottky Varactor Diode Millimeter-Wave Frequency Multipliers

DTIC Science & Technology

1992-06-01

wave applications", IEEE Trans on Microwave Theory and Tech., vol. 39, no. 12, Dec. 1991 , pp. 1964-1971. A copy of this paper is 35 included in...Watts to Bulky 1991 spectral HV DC Power line Pwr Very Inguscio varies Massive 1986 with Vac.:um line Very low Gas noise Supply Ledatron Up to 1 W at...PULSED Band up to 1985 HV DC 10 GHz Massive Pwr Magnetic V?4MA > 100 GHz > 1 Watt Wide Cooling Research Quasi- McGruer Theory Theory Band Planar 1991

Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Billone, M. C.; Burtseva, T. A.

2016-08-30

The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

Feasibility of recycling thorium in a fusion-fission hybrid/PWR symbiotic system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Josephs, J.M.

1980-12-31

A study was made of the economic impact of high levels of radioactivity in the thorium fuel cycle. The sources of this radioactivity and means of calculating the radioactive levels at various stages in the fuel cycle are discussed and estimates of expected levels are given. The feasibility of various methods of recycling thorium is discussed. These methods include direct recycle, recycle after storage for 14 years to allow radioactivity to decrease, shortening irradiation times to limit radioactivity build up, and the use of the window in time immediately after reprocessing where radioactivity levels are diminished. An economic comparison ismore » made for the first two methods together with the throwaway option where thorium is not recycled using a mass energy flow model developed for a CTHR (Commercial Tokamak Hybrid Reactor), a fusion fission hybrid reactor which serves as fuel producer for several PWR reactors. The storage option is found to be most favorable; however, even this option represents a significant economic impact due to radioactivity of 0.074 mills/kW-h which amounts to $4 x 10/sup 9/ over a 30 year period assuming a 200 gigawatt supply of electrical power.« less

Fundamental metallurgical aspects of axial splitting in zircaloy cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, H. M.

Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10{sup 21} n cm{sup {minus}2} to 5.9 x 10{sup 21} n cm{sup {minus}2} (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest claddingmore » were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed.« less

BNL severe-accident sequence experiments and analysis program. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, G.A.; Ginsberg, T.; Tutu, N.K.

1983-01-01

In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

NASA Astrophysics Data System (ADS)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Preliminary Stratigraphic Basis for Geologic Mapping of Venus

NASA Technical Reports Server (NTRS)

Basilevsky, A. T.; Head, J. W.

1993-01-01

The age relations between geologic formations have been studied at 36 1000x1000 km areas centered at the dark paraboloid craters. The geologic setting in all these sites could be characterized using only 16 types of features and terrains (units). These units form a basic stratigraphic sequence (from older to younger: (1) Tessera (Tt); (2-3) Densely fractured terrains associated with coronae (COdf) and in the form of remnants among plains (Pdf); (4) Fractured and ridged plains (Pfr); (5) Plains with wrinkle ridges (Pwr); (6-7) Smooth and lobate plains (Ps/Pl); and (8) Rift-associated fractures (Fra). The stratigraphic position of the other units is determined by their relation with the units of the basic sequence: (9) Ridge bells (RB), contemporary with Pfr; (10-11) Ridges of coronae and arachnoids annuli (COar/Aar), contemporary with wrinkle ridges of Pwr; (12) Fractures of coronae annuli (COaf) disrupt Pwr and Ps/Pl; (13) Fractures (F) disrupt Pwr or younger units; (14) Craters with associated dark paraboloids (Cdp), which are on top of all volcanic and tectonic units except the youngest episodes of rift-associated fracturing and volcanism; (15-16) Surficial streaks (Ss) and surficial patches (Sp) are approximately contemporary with Cdp. These units may be used as a tentative basis for the geologic mapping of Venus including VMAP. This mapping should test the stratigraphy and answer the question of whether this stratigraphic sequence corresponds to geologic events which were generally synchronous all around the planet or whether the sequence is simply a typical sequence of events which occurred in different places at diffferent times.

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Coexistence of insulin resistance and increased glucose tolerance in pregnant rats: a physiological mechanism for glucose maintenance.

PubMed

Carrara, Marcia Aparecida; Batista, Márcia Regina; Saruhashi, Tiago Ribeiro; Felisberto, Antonio Machado; Guilhermetti, Marcio; Bazotte, Roberto Barbosa

2012-06-06

The contribution of insulin resistance (IR) and glucose tolerance to the maintenance of blood glucose levels in non diabetic pregnant Wistar rats (PWR) was investigated. PWR were submitted to conventional insulin tolerance test (ITT) and glucose tolerance test (GTT) using blood sample collected 0, 10 and 60 min after intraperitoneal insulin (1 U/kg) or oral (gavage) glucose (1g/kg) administration. Moreover, ITT, GTT and the kinetics of glucose concentration changes in the fed and fasted states were evaluated with a real-time continuous glucose monitoring system (RT-CGMS) technique. Furthermore, the contribution of the liver glucose production was investigated. Conventional ITT and GTT at 0, 7, 14 and 20 days of pregnancy revealed increased IR and glucose tolerance after 20 days of pregnancy. Thus, this period of pregnancy was used to investigate the kinetics of glucose changes with the RT-CGMS technique. PWR (day 20) exhibited a lower (p<0.05) glucose concentration in the fed state. In addition, we observed IR and increased glucose tolerance in the fed state (PWR-day 20 vs. day 0). Furthermore, our data from glycogenolysis and gluconeogenesis suggested that the liver glucose production did not contribute to these changes in insulin sensitivity and/or glucose tolerance during late pregnancy. In contrast to the general view that IR is a pathological process associated with gestational diabetes, a certain degree of IR may represent an important physiological mechanism for blood glucose maintenance during fasting. Copyright © 2012 Elsevier Inc. All rights reserved.

The pluralistic water research concept - a new human-water system research approach

NASA Astrophysics Data System (ADS)

Evers, Mariele; Höllermann, Britta; Almoradie, Adrian; Taft, Linda; Garcia-Santos, Glenda

2017-04-01

Sustainable water resources management has been and still is a main challenge for decision makers even though for the past number of decades integrative approaches and concepts (e.g. Integrated Water Resources Management - IWRM) have been developed to address problems on floods, droughts, water quality, water quantity, environment and ecology. Although somehow these approaches are aiming to address water related problems in an integrative approach and to some extent include or involve society in the planning and management, they still lack some of the vital components in including the social dimensions and their interaction with water. Understanding these dynamics in a holistic way and how they are shaped by time and space may tackle these shortcomings and provide more effective and sustainable management solutions with respect to a set of potential present social actions and values as well as possible futures. This paper aims to discuss challenges to coherently and comprehensively integrate the social dimensions of different human-water concepts like IWRM, socio-hydrology and waterscape. Against this background it will develop criteria for an integrative approach and present a newly developed concept termed pluralistic water research (PWR) concept. PWR is not only a pluralistic but also an integrative and interdisciplinary approach to acknowledge the social and water dimensions and their interaction and dynamics by considering more than one perspective of a water-related issue, hereby providing a set of multiple (future) developments. Our PWR concept will be illustrated by a case study application of the Canary island La Gomera. Furthermore an outlook on further possible developments of the PWR concept will be presented and discussed.

Fabrication of simulated DUPIC fuel

NASA Astrophysics Data System (ADS)

Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

2000-12-01

Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

Development of a new lattice physics code robin for PWR application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, S.; Chen, G.

2013-07-01

This paper presents a description of methodologies and preliminary verification results of a new lattice physics code ROBIN, being developed for PWR application at Shanghai NuStar Nuclear Power Technology Co., Ltd. The methods used in ROBIN to fulfill various tasks of lattice physics analysis are an integration of historical methods and new methods that came into being very recently. Not only these methods like equivalence theory for resonance treatment and method of characteristics for neutron transport calculation are adopted, as they are applied in many of today's production-level LWR lattice codes, but also very useful new methods like the enhancedmore » neutron current method for Dancoff correction in large and complicated geometry and the log linear rate constant power depletion method for Gd-bearing fuel are implemented in the code. A small sample of verification results are provided to illustrate the type of accuracy achievable using ROBIN. It is demonstrated that ROBIN is capable of satisfying most of the needs for PWR lattice analysis and has the potential to become a production quality code in the future. (authors)« less

Regeneratively Cooled Liquid Oxygen/Methane Technology Development Between NASA MSFC and PWR

NASA Technical Reports Server (NTRS)

Robinson, Joel W.; Greene, Christopher B.; Stout, Jeffrey B.

2012-01-01

The National Aeronautics & Space Administration (NASA) has identified Liquid Oxygen (LOX)/Liquid Methane (LCH4) as a potential propellant combination for future space vehicles based upon exploration studies. The technology is estimated to have higher performance and lower overall systems mass compared to existing hypergolic propulsion systems. NASA-Marshall Space Flight Center (MSFC) in concert with industry partner Pratt & Whitney Rocketdyne (PWR) utilized a Space Act Agreement to test an oxygen/methane engine system in the Summer of 2010. PWR provided a 5,500 lbf (24,465 N) LOX/LCH4 regenerative cycle engine to demonstrate advanced thrust chamber assembly hardware and to evaluate the performance characteristics of the system. The chamber designs offered alternatives to traditional regenerative engine designs with improvements in cost and/or performance. MSFC provided the test stand, consumables and test personnel. The hot fire testing explored the effective cooling of one of the thrust chamber designs along with determining the combustion efficiency with variations of pressure and mixture ratio. The paper will summarize the status of these efforts.

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

Fuel thermal conductivity (FTHCON). Status report. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagrman, D. L.

1979-02-01

An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on amore » more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced.« less

Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

NASA Astrophysics Data System (ADS)

Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

2016-04-01

This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

NASA Astrophysics Data System (ADS)

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurindranath; Natesan, Krishnamurti

2016-05-01

Argonne National Laboratory (ANL), under the sponsorship of Department of Energy's Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316 SS) material which is widely used in the US reactors. Contrary to the conventional S ∼ N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. Mechanics-based modeling of fatigue such as by using finite element (FE) tools requires the time/cycle dependent material hardening properties. Presently such time-dependent material hardening properties are hardly available in fatigue modeling literature even under in-air conditions. Getting those material properties under PWR environment, are even harder. Through this work we made preliminary attempt to generate time/cycle dependent stress-strain data both under in-air and PWR water conditions for further study such as for possible development of material models and constitutive relations for FE model implementation. Although, there are open-ended possibility to further improve the discussed test methods and related material estimation techniques we anticipate that the data presented in this paper will help the metal fatigue research community particularly, the researchers who are dealing with mechanistic modeling of metal fatigue such as using FE tools. In this paper the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed.

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Experimental evaluation of models for predicting Cherenkov light intensities from short-cooled nuclear fuel assemblies

NASA Astrophysics Data System (ADS)

Branger, E.; Grape, S.; Jansson, P.; Jacobsson Svärd, S.

2018-02-01

The Digital Cherenkov Viewing Device (DCVD) is a tool used by nuclear safeguards inspectors to verify irradiated nuclear fuel assemblies in wet storage based on the recording of Cherenkov light produced by the assemblies. One type of verification involves comparing the measured light intensity from an assembly with a predicted intensity, based on assembly declarations. Crucial for such analyses is the performance of the prediction model used, and recently new modelling methods have been introduced to allow for enhanced prediction capabilities by taking the irradiation history into account, and by including the cross-talk radiation from neighbouring assemblies in the predictions. In this work, the performance of three models for Cherenkov-light intensity prediction is evaluated by applying them to a set of short-cooled PWR 17x17 assemblies for which experimental DCVD measurements and operator-declared irradiation data was available; (1) a two-parameter model, based on total burnup and cooling time, previously used by the safeguards inspectors, (2) a newly introduced gamma-spectrum-based model, which incorporates cycle-wise burnup histories, and (3) the latter gamma-spectrum-based model with the addition to account for contributions from neighbouring assemblies. The results show that the two gamma-spectrum-based models provide significantly higher precision for the measured inventory compared to the two-parameter model, lowering the standard deviation between relative measured and predicted intensities from 15.2 % to 8.1 % respectively 7.8 %. The results show some systematic differences between assemblies of different designs (produced by different manufacturers) in spite of their similar PWR 17x17 geometries, and possible ways are discussed to address such differences, which may allow for even higher prediction capabilities. Still, it is concluded that the gamma-spectrum-based models enable confident verification of the fuel assembly inventory at the currently used detection limit for partial defects, being a 30 % discrepancy between measured and predicted intensities, while some false detection occurs with the two-parameter model. The results also indicate that the gamma-spectrum-based prediction methods are accurate enough that the 30 % discrepancy limit could potentially be lowered.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leskovar, Matjaz; Koncar, Bostjan

An ex-vessel steam explosion may occur when during a severe reactor accident the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles at later times, during the expansion of the highly pressurized water vapor, that may endanger surrounding structures. In contrast to specialized steammore » explosion CFD codes, where the steam explosion is modeled on micro-scale using fundamental averaged multiphase flow conservation equations, in the presented approach the steam explosion is modeled in a simplified manner as an expanding high-pressure pre-mixture of dispersed molten fuel, liquid water and vapor. Applying the developed steam explosion model, a comprehensive analysis of the ex-vessel steam explosion in a typical PWR reactor cavity was done using the CFD code CFX-10. At four selected locations, which are of importance for the assessment of the vulnerability of cavity structures, the pressure histories were recorded and the corresponding pressure impulses calculated. The pressure impulses determine the destructive potential of the steam explosion and represent the input for the structural mechanical analysis of the cavity structures. The simulation results show that the pressure impulses depend mainly on the steam explosion energy conversion ratio, whereas the influence of the pre-mixture vapor volume fraction, which is a parameter in our model and determines the maximum steam explosion pressure, is not significant. (authors)« less

Xenon-induced power oscillations in a generic small modular reactor

NASA Astrophysics Data System (ADS)

Kitcher, Evans Damenortey

As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a preliminary investigation into excess reactivity control options for the SMR design was conducted confirming the generally held notion that existing PWR control mechanisms can be used in iPWR SMRs with similar effectiveness. With the desire to operate the SMR under the boron free coolant condition, erbium oxide fuel integral burnable absorber rods were identified as a possible means to retain the dispersed absorber effect of soluble boron in the reactor coolant in replacement.

Posttest calculation of the PBF LOC-11B and LOC-11C experiments using RELAP4/MOD6. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hendrix, C.E.

Comparisons between RELAP4/MOD6, Update 4 code-calculated and measured experimental data are presented for the PBF LOC-11C and LOC-11B experiments. Independent code verification techniques are now being developed and this study represents a preliminary effort applying structured criteria for developing computer models, selecting code input, and performing base-run analyses. Where deficiencies are indicated in the base-case representation of the experiment, methods of code and criteria improvement are developed and appropriate recommendations are made.

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Szilard, Ronaldo Henriques

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

75 FR 13 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

...The Nuclear Regulatory Commission (NRC) is amending its regulations to provide alternate fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. This final rule provides alternate PTS requirements based on updated analysis methods. This action is desirable because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. This action reduces regulatory burden for those PWR licensees who expect to exceed the existing requirements before the expiration of their licenses, while maintaining adequate safety, and may choose to comply with the final rule as an alternative to complying with the existing requirements.

Performance testing and analyses of the VSC-17 ventilated concrete cask. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-05-01

This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power`s Surry and Florida Power & Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained.

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Bernard, D.; Blaise, P.

2013-07-01

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency ofmore » Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

OLIGOCELLULA1/HIGH EXPRESSION OF OSMOTICALLY RESPONSIVE GENES15 Promotes Cell Proliferation With HISTONE DEACETYLASE9 and POWERDRESS During Leaf Development in Arabidopsis thaliana

PubMed Central

Suzuki, Marina; Shinozuka, Nanae; Hirakata, Tomohiro; Nakata, Miyuki T.; Demura, Taku; Tsukaya, Hirokazu; Horiguchi, Gorou

2018-01-01

Organ size regulation is dependent on the precise spatial and temporal regulation of cell proliferation and cell expansion. A number of transcription factors have been identified that play a key role in the determination of aerial lateral organ size, but their functional relationship to various chromatin modifiers has not been well understood. To understand how leaf size is regulated, we previously isolated the oligocellula1 (oli1) mutant of Arabidopsis thaliana that develops smaller first leaves than the wild type (WT) mainly due to a reduction in the cell number. In this study, we further characterized oli1 leaf phenotypes and identified the OLI1 gene as well as interaction partners of OLI1. Detailed characterizations of leaf development suggested that the cell proliferation rate in oli1 leaf primordia is lower than that in the WT. In addition, oli1 was associated with a slight delay of the progression from the juvenile to adult phases of leaf traits. A classical map-based approach demonstrated that OLI1 is identical to HIGH EXPRESSION OF OSMOTICALLY RESPONSIVE GENES15 (HOS15). HOS15/OLI1 encodes a homolog of human transducin β-like protein1 (TBL1). TBL1 forms a transcriptional repression complex with the histone deacetylase (HDAC) HDAC3 and either nuclear receptor co-repressor (N-CoR) or silencing mediator for retinoic acid and thyroid receptor (SMRT). We found that mutations in HISTONE DEACETYLASE9 (HDA9) and a switching-defective protein 3, adaptor 2, N-CoR, and transcription factor IIIB-domain protein gene, POWERDRESS (PWR), showed a small-leaf phenotype similar to oli1. In addition, hda9 and pwr did not further enhance the oli1 small-leaf phenotype, suggesting that these three genes act in the same pathway. Yeast two-hybrid assays suggested physical interactions, wherein PWR probably bridges HOS15/OLI1 and HDA9. Earlier studies suggested the roles of HOS15, HDA9, and PWR in transcriptional repression. Consistently, transcriptome analyses showed several genes commonly upregulated in the three mutants. From these findings, we propose a possibility that HOS15/OLI1, PWR, and HDA9 form an evolutionary conserved transcription repression complex that plays a positive role in the regulation of final leaf size. PMID:29774040

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

NASA Astrophysics Data System (ADS)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Hybrid parallel code acceleration methods in full-core reactor physics calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Courau, T.; Plagne, L.; Ponicot, A.

2012-07-01

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadraturemore » required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)« less

ADVANCEMENTS IN TIME-SPECTRA ANALYSIS METHODS FOR LEAD SLOWING-DOWN SPECTROSCOPY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Leon E.; Anderson, Kevin K.; Gesh, Christopher J.

2010-08-11

Direct measurement of Pu in spent nuclear fuel remains a key challenge for safeguarding nuclear fuel cycles of today and tomorrow. Lead slowing-down spectroscopy (LSDS) is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic mass with an uncertainty lower than the approximately 10 percent typical of today’s confirmatory assay methods. Pacific Northwest National Laboratory’s (PNNL) previous work to assess the viability of LSDS for the assay of pressurized water reactor (PWR) assemblies indicated that the method could provide direct assay of Pu-239 and U-235 (and possibly Pu-240 and Pu-241)more » with uncertainties less than a few percent, assuming suitably efficient instrumentation, an intense pulsed neutron source, and improvements in the time-spectra analysis methods used to extract isotopic information from a complex LSDS signal. This previous simulation-based evaluation used relatively simple PWR fuel assembly definitions (e.g. constant burnup across the assembly) and a constant initial enrichment and cooling time. The time-spectra analysis method was founded on a preliminary analytical model of self-shielding intended to correct for assay-signal nonlinearities introduced by attenuation of the interrogating neutron flux within the assembly.« less

SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandiamore » National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The cases with two-phase flow at the turbine inlet will be pursued in future work.« less

Zebra: An advanced PWR lattice code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, L.; Wu, H.; Zheng, Y.

2012-07-01

This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precisionmore » and a high efficiency. (authors)« less

Estimating probable flaw distributions in PWR steam generator tubes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorman, J.A.; Turner, A.P.L.

1997-02-01

This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regardingmore » uncertainties and assumptions in the data and analyses.« less

Thermal modeling of a vertical dry storage cask for used nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Jie; Liu, Yung Y.

2016-05-01

Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17 x 17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. Themore » effects of canister fill gas (helium or nitrogen), internal pressure (1-6 atm), and basket material (stainless steel or aluminum alloy) were studied to determine the peak cladding temperature (PCT) and the canister surface temperatures (CSTs). The results showed that high thermal conductivity of the basket material greatly enhances heat transfer and reduces the PCT. The results also showed that natural convection affects both PCT and the CST profile, while the latter depends strongly on the type of fill gas and canister internal pressure. Of particular interest to condition and performance monitoring is the identification of canister locations where significant temperature change occurs after a canister is breached and the fill gas changes from high-pressure helium to ambient air. This study provided insight on the thermal performance of a vertical storage cask containing high-burnup fuel, and helped advance the concept of monitoring CSTs as a means to detect helium leakage from a welded canister. The effects of blockage of air inlet vents on the cask's thermal performance were studied. The simulation were validated by comparing the results against data obtained from the temperature measurements of a commercial cask.« less

Analysis and Implementation of Accident Tolerant Nuclear Fuels

NASA Astrophysics Data System (ADS)

Prewitt, Benjamin Joseph

To improve the reliability and robustness of LWR, accident tolerant nuclear fuels and cladding materials are being developed to possibly replace the current UO2/zirconium system. This research highlights UN and U3Si 2, two of the most favorable accident tolerant fuels being developed. To evaluate the commercial feasiblilty of these fuels, two areas of research were conducted. Chemical fabrication routes for both fuels were investigated in detail, considering UO2 and UF6 as potential starting materials. Potential pathways for industrial scale fabrication using these methods were discussed. Neutronic performance of 70%UN-30%U3Si2 composite was evaluated in MNCP using PWR assembly and core models. The results showed comparable performance to an identical UO2 fueled simulation with the same configuration. The parameters simulated for composite and oxide fuel include the following: fuel to moderator ratio curves; energy dependent flux spectra; temperature coefficients for fuel and moderator; delayed neutron fractions; power peaking factors; axial and radial flux profiles in 2D and 3D; burnup; critical boron concentration; and shutdown margin. Overall, the neutronic parameters suggest that the transition from UO2 to composite in existing nuclear systems will not require significant changes in operating procedures or modifications to standards and regulations.

Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.

Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent researchmore » progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.« less

Human Factors and Technical Considerations for a Computerized Operator Support System Prototype

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ulrich, Thomas Anthony; Lew, Roger Thomas; Medema, Heather Dawne

2015-09-01

A prototype computerized operator support system (COSS) has been developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Departmentmore » of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment. A COSS demonstration scenario has been developed for the prototype involving the Chemical & Volume Control System (CVCS) of the PWR simulator. It involves a primary coolant leak outside of containment that would require tripping the reactor if not mitigated in a very short timeframe. The COSS prototype presents a series of operator screens that provide the needed information and soft controls to successfully mitigate the event.« less

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2011-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

Research gaps and technology needs in development of PHM for passive AdvSMR components

NASA Astrophysics Data System (ADS)

Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Hirt, Evelyn H.; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henagar, Chuck H., Jr.

2014-02-01

Advanced small modular reactors (AdvSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near-term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. SMRs are challenged economically because of losses in economy of scale; thus, there is increased motivation to reduce the controllable operations and maintenance costs through automation technologies including prognostics health management (PHM) systems. In this regard, PHM systems have the potential to play a vital role in supporting the deployment of AdvSMRs and face several unique challenges with respect to implementation for passive AdvSMR components. This paper presents a summary of a research gaps and technical needs assessment performed for implementation of PHM for passive AdvSMR components.

Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fridman, E.; Shwageraus, E.; Galperin, A.

2006-07-01

The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of thismore » work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)« less

Analysis of the Geologic Structure and Compilation of the Geologic Map of the Northern Part of Planet Venus

NASA Astrophysics Data System (ADS)

Basilevsky, A. T.; Burba, G. A.; Ivanov, M. A.; Bobina, N. N.; Shashkina, V. P.; Head, J. W.

Based on an analysis of the images of the Venusian surface obtained by the side-looking radar of the Magellan orbiter, a geologic map of the northern part of Venus (the region extending to the north of the 35°N latitude) at 1 : 10 000 000 scale is compiled. The map of this vast territory, comprising one-fifth of the planet surface, was compiled using only 12 geologic units, which implies a uniform character of terrains and land- forms on the investigated territory and, therefore, the uniformity of geologic processes that occurred on this planet. These units are the products of four main groups of geologic processes that occurred on Venus during the last 0.51 Myr: (1) basaltic volcanism; (2) tectonic compression and tensile deformation; (3) impact crater- ing; and (4) wind-related mobilization, transportation, and deposition of loose fine-grained materials. Basaltic volcanism is the main process that supplies new material on the surface of Venus. Tectonic deformation struc- tures, superposed on the material of different geologic units, determined the morphology of the units and formed the surfaces of unconformity between neighboring units. Ten of 12 geologic units form an age sequence that is virtually identical over the entire mapped territory of the planet. The possible incon- sistency of this sequence caused by anomalous relations existing between smooth plains (Ps) in the southeastern part of Lakshmi Planum and wrinkle ridged plains (Pwr) in the northern part of Sedna Planitia does not destroy this sequence as a whole. The results of our mapping support the model of global stratigraphy of Venus proposed by Basilevsky and Head (19951998) and provide evidence of the quasi-synchronous character of single-type geologic units on different areas of Venus rather than of the absence of synchronism. An analysis of the distribution of impact craters on different geologic units has shown the proximity of mean absolute ages of the material of the surface of Pwr plains, of the entire studied territory, and of the entire Venusian surface. The results of our analysis suggest that, within the area under study, the intensity of the leading geologic processes at the beginning of the studied segment of the geologic history was relatively high but decreased dramatically later.

K-TIF: a two-fluid computer program for downcomer flow dynamics. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amsden, A.A.; Harlow, F.H.

1977-10-01

The K-TIF computer program has been developed for numerical solution of the time-varying dynamics of steam and water in a pressurized water reactor downcomer. The current status of physical and mathematical modeling is presented in detail. The report also contains a complete description of the numerical solution technique, a full description and listing of the computer program, instructions for its use, with a sample printout for a specific test problem. A series of calculations, performed with no change in the modeling parameters, shows consistent agreement with the experimental trends over a wide range of conditions, which gives confidence to themore » calculations as a basis for investigating the complicated physics of steam-water flows in the downcomer.« less

Influence of Stress Corrosion Crack Morphology on Ultrasonic Examination Performances

NASA Astrophysics Data System (ADS)

Dupond, O.; Duwig, V.; Fouquet, T.

2009-03-01

Stress Corrosion Cracking represents a potential damage for several components in PWR. For this reason, NDE of stress corrosion cracks corresponds to an important stake for Electricité de France (EDF) both for availability and for safety of plants. This paper is dedicated to the ultrasonic examination of SCC crack defects. The study mixes an experimental approach conducted on artificial flaws—meant to represent the characteristic morphologic features often encountered on SCC cracks—and a 2D finite element modelling with the code ATHENA 2D developed by EDF. Results indicate that ATHENA reproduces correctly the interaction of the beam on the complex defect. Indeed specific ultrasonic responses resulting from the defect morphology have been observed experimentally and reproduced with the modelling.

High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

2014-08-01

In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover, for as high as 1600 °C, the non-reactive LM was experimentally confirmed not to show any chemical reaction with air or moisture in the air. This experimental work confirmed the excellent compatibility behaviors between the LM as a PWR fuel gap filler and stainless steel and high-density concrete in the high-temperature regime.

Advanced Energy Harvesting Control Schemes for Marine Renewable Energy Devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

McEntee, Jarlath; Polagye, Brian; Fabien, Brian

2016-03-31

The Advanced Energy Harvesting Control Schemes for Marine Renewable Energy Devices (Project) investigated, analyzed and modeled advanced turbine control schemes with the objective of increasing the energy harvested by hydrokinetic turbines in turbulent flow. Ocean Renewable Power Company (ORPC) implemented and validated a feedforward controller to increase power capture; and applied and tested the controls on ORPC’s RivGen® Power Systems in Igiugig, Alaska. Assessments of performance improvements were made for the RivGen® in the Igiugig environment and for ORPC’s TidGen® Power System in a reference tidal environment. Annualized Energy Production (AEP) and Levelized Cost of Energy (LCOE) improvements associated withmore » implementation of the recommended control methodology were made for the TidGen® Power System in the DOE reference tidal environment. System Performance Advancement (SPA) goals were selected for the project. SPA targets were to improve Power to Weight Ratio (PWR) and system Availability, with the intention of reducing Levelized Cost of Electricity (LCOE). This project focused primarily reducing in PWR. Reductions in PWR of 25.5% were achieved. Reductions of 20.3% in LCOE were achieved. This project evaluated four types of controllers which were tested in simulation, emulation, a laboratory flume, and the field. The adaptive Kω2 controller performs similarly to the non-adaptive version of the same controller and may be useful in tidal channels where the mean velocity is continually evolving. Trends in simulation were largely verified through experiments, which also provided the opportunity to test assumptions about turbine responsiveness and control resilience to varying scales of turbulence. Laboratory experiments provided an essential stepping stone between simulation and implementation on a field-scale turbine. Experiments also demonstrated that using “energy loss” as a metric to differentiate between well-designed controllers operating at an optimal tip-speed ratio set-point is difficult, which anticipated the outcome from field experiments. The clear message is that the feedforward Kω2 controller out-performs the feedback controllers in almost all aspects and modes of evaluation. The controllers proved a substantial improvement over the baseline performance of the TidGen® turbine, in terms of energy capture. The effects of noise-contaminated angular velocity signals were investigated and validated by simulation as an explanation for the performance limitations observed for TidGen® turbine operations in Eastport, Maine. Measurements of loads performed as part of the laboratory testing indicate that there are limited differences in average axial thrust force between control architectures. This suggests that none of the control strategies are likely to substantially affect loads on the turbine support structure. Velocity measurements during the ORPC RivGen® turbine deployment at Igiugig, Alaska, in 2014 were used to assess the variability of the river flow. Results suggest that the river flow is approximately steady, in the mean sense, at any particular location in the river, with random turbulent fluctuations that are around 10% of the mean flow. The mean flow in the center channel of the river is 2.5 m/s, with reductions near the riverbanks and in the shallows. Spectral analysis and lagged correlation results indicate that temporal fluctuations at a given point are dominated by large scale fluctuations, such that measurements at the turbine location are just as useful for inflow control implementation as upstream measurements. At this site, and likely at many other river sites, flow is generally steady at a given location, but flow varies dramatically between locations, particularly laterally across the river. The primary result is that a lateral change in position of a few meters results in changes to flow speed that far exceed the turbulence fluctuations at any given location. The turbulence is dominated by long time scales. Following final system tests, the RivGen® device was submerged and each evaluated controller was tested across a range of gain/set point values and filter configurations for a minimum of 5 minutes, with longer runs attempted for well-performing cases. In addition to testing controllers during the 2015 deployment season, LGL Alaska Research Associates Inc. (LGL) performed a fish monitoring study in compliance with Alaska Department of Fish and Game fisheries habitat permit for the project. During the season, LGL reviewed 10% of 179 one-hour blocks of footage (6 minutes on the hour) and documented a total of over 1200 fish in the vicinity of the RivGen® device, including over 800 salmon smolt and over 350 adult salmon. No evidence of adverse effects including passage delay by upstream migrating salmon was noted. This is an important result for future deployments and has a direct impact on commercial system designs.« less

Efficiency, equity and feasibility of strategies to identify the poor: an application to premium exemptions under National Health Insurance in Ghana.

PubMed

Jehu-Appiah, Caroline; Aryeetey, Genevieve; Spaan, Ernst; Agyepong, Irene; Baltussen, Rob

2010-05-01

This paper outlines the potential strategies to identify the poor, and assesses their feasibility, efficiency and equity. Analyses are illustrated for the case of premium exemptions under National Health Insurance (NHI) in Ghana. A literature search in Medline search was performed to identify strategies to identify the poor. Models were developed including information on demography and poverty, and costs and errors of in- and exclusion of these strategies in two regions in Ghana. Proxy means testing (PMT), participatory welfare ranking (PWR), and geographic targeting (GT) are potentially useful strategies to identify the poor, and vary in terms of their efficiency, equity and feasibility. Costs to exempt one poor individual range between US$11.63 and US$66.67, and strategies may exclude up to 25% of the poor. Feasibility of strategies is dependent on their aptness in rural/urban settings, and administrative capacity to implement. A decision framework summarizes the above information to guide policy making. We recommend PMT as an optimal strategy in relative low poverty incidence urbanized settings, PWR as an optimal strategy in relative low poverty incidence rural settings, and GT as an optimal strategy in high incidence poverty settings. This paper holds important lessons not only for NHI in Ghana but also for other countries implementing exemption policies. Copyright (c) 2009 Elsevier Ireland Ltd. All rights reserved.

Analysis of steam generator tube rupture transients with single failure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

The Gesellschaft fuer Reaktorsicherheit is engaged in the collection and evaluation of light water reactor operating experience as well as analyses for the risk study of the pressurized water reactor (PWR). Within these activities, thermohydraulic calculations have been performed to show the influence of different boundary conditions and disturbances on the steam generator tube rupture (SGTR) transients. The analyses of these calculations have focused on the measures and systems needed to cope with an SGTR. The reference plant for this analysis is a 1300-MW(e) PWR of Kraftwerk Union design with four loops, each containing a U-tube steam generator (SG) andmore » a reactor cooling pump (RCP). The thermal-hydraulic code DRUFAN-02 was used for the transient calculations.« less

SCC Initiation Behavior of Alloy 182 in PWR Primary Water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toloczko, Mychailo B.; Zhai, Ziqing; Bruemmer, Stephen M.

SCC initiation behavior of 15% cold forged specimens cut from four different alloy 182 weldments was investigated in 360°C simulated PWR primary water under constant load at the yield stress using direct current potential drop to perform in-situ monitoring of SCC initiation time. Within each weldment, one or more specimens underwent SCC initiation within 24 hours of reaching full load while some specimens had much longer initiation times, in a few cases exceeding 2500 hours. Detailed examinations were conducted on these specimens with a focus on different microstructural features such as preexisting defects, grain orientation and second phases, highlighting anmore » important role of microstructure in crack initiation of alloy 182.« less

Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

De Rosa, Felice

2006-07-01

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to themore » accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its specific models (candling, corium pool behaviour, etc.) they were less good. A future work will be the preparation of an input deck for the new MELCOR 1.8.6. and to perform a code-to-code comparison with ASTEC v1.2 rev. 1. (author)« less

Contribution to the thermodynamic description of the corium - The U-Zr-O system

NASA Astrophysics Data System (ADS)

Quaini, A.; Guéneau, C.; Gossé, S.; Dupin, N.; Sundman, B.; Brackx, E.; Domenger, R.; Kurata, M.; Hodaj, F.

2018-04-01

In order to understand the stratification process that may occur in the late phase of the fuel degradation during a severe accident in a PWR, the thermodynamic knowledge of the U-Zr-O system is crucial. The presence of a miscibility gap in the U-Zr-O liquid phase may lead to a stratified configuration, which will impact the accidental scenario management. The aim of this work was to obtain new experimental data in the U-Zr-O liquid miscibility gap. New tie-line data were provided at 2567 ± 25 K. The related thermodynamic models were reassessed using present data and literature values. The reassessed model will be implemented in the TAF-ID international database. The composition and density of phases potentially formed during stratification will be predicted by coupling current thermodynamic model with thermal-hydraulics codes.

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.« less

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

76 FR 41783 - Combined Notice of Filings #2

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-15

... Commodities Group, Constellation Pwr Source Generation LLC, Constellation NewEnergy, Inc., CER Generation II..., CER Generation, LLC, Constellation Energy Commodities Group M, Constellation Mystic Power, LLC...

Experimental validation of the DARWIN2.3 package for fuel cycle applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

San-Felice, L.; Eschbach, R.; Bourdot, P.

2012-07-01

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuelmore » inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)« less

Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

2012-07-01

The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existingmore » facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)« less

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forget, Benoit; Smith, Kord; Kumar, Shikhar

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevantmore » multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.« less

Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

NASA Astrophysics Data System (ADS)

Khuwaileh, Bassam

High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL) based algorithm previously developed to quantify the uncertainty for single physics models is extended for large scale multi-physics coupled problems with feedback effect. Moreover, a non-linear surrogate based UQ approach is developed, used and compared to performance of the KL approach and brute force Monte Carlo (MC) approach. On the other hand, an efficient Data Assimilation (DA) algorithm is developed to assess information about model's parameters: nuclear data cross-sections and thermal-hydraulics parameters. Two improvements are introduced in order to perform DA on the high dimensional problems. First, a goal-oriented surrogate model can be used to replace the original models in the depletion sequence (MPACT -- COBRA-TF - ORIGEN). Second, approximating the complex and high dimensional solution space with a lower dimensional subspace makes the sampling process necessary for DA possible for high dimensional problems. Moreover, safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. Accordingly, an inverse problem can be defined and solved to assess the contributions from sources of uncertainty; and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this dissertation a subspace-based gradient-free and nonlinear algorithm for inverse uncertainty quantification namely the Target Accuracy Assessment (TAA) has been developed and tested. The ideas proposed in this dissertation were first validated using lattice physics applications simulated using SCALE6.1 package (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lattice models). Ultimately, the algorithms proposed her were applied to perform UQ and DA for assembly level (CASL progression problem number 6) and core wide problems representing Watts Bar Nuclear 1 (WBN1) for cycle 1 of depletion (CASL Progression Problem Number 9) modeled via simulated using VERA-CS which consists of several multi-physics coupled models. The analysis and algorithms developed in this dissertation were encoded and implemented in a newly developed tool kit algorithms for Reduced Order Modeling based Uncertainty/Sensitivity Estimator (ROMUSE).

PWR design for low doses in the United Kingdom: The present and the future

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zodiates, A.M.; Willcock, A.

1995-03-01

The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B,more » presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.« less

Development of on-line monitoring system for Nuclear Power Plant (NPP) using neuro-expert, noise analysis, and modified neural networks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subekti, M.; Center for Development of Reactor Safety Technology, National Nuclear Energy Agency of Indonesia, Puspiptek Complex BO.80, Serpong-Tangerang, 15340; Ohno, T.

2006-07-01

The neuro-expert has been utilized in previous monitoring-system research of Pressure Water Reactor (PWR). The research improved the monitoring system by utilizing neuro-expert, conventional noise analysis and modified neural networks for capability extension. The parallel method applications required distributed architecture of computer-network for performing real-time tasks. The research aimed to improve the previous monitoring system, which could detect sensor degradation, and to perform the monitoring demonstration in High Temperature Engineering Tested Reactor (HTTR). The developing monitoring system based on some methods that have been tested using the data from online PWR simulator, as well as RSG-GAS (30 MW research reactormore » in Indonesia), will be applied in HTTR for more complex monitoring. (authors)« less

Electrochemical study of pre- and post-transition corrosion of Zr alloys in PWR coolant

NASA Astrophysics Data System (ADS)

Macák, Jan; Novotný, Radek; Sajdl, Petr; Renčiuková, Veronika; Vrtílková, Věra

Corrosion properties of Zr-Sn and Zr-Nb zirconium alloys were studied under simulated PWR conditions (or, more exactly, VVER conditions — boric acid, potassium hydroxide, lithium hydroxide) at temperatures up to 340°C and 15MPa using in-situ electrochemical impedance spectroscopy (EIS) and polarization measurements. EIS spectra were obtained in a wide range of frequencies (typically 100kHz — 100μHz). It enabled to gain information of both dielectric properties of oxide layers developing on the Zr-alloys surface and of the kinetics of the corrosion process and the associated charge and mass transfer phenomena. Experiments were run for more than 380 days; thus, the study of all the corrosion stages (pre-transition, transition, post-transition) was possible.

Conceptual design study of small long-life PWR based on thorium cycle fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul

2014-09-30

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWRmore » result small excess reactivity and reduced power peaking during its operation.« less

Efficient solution of the simplified P N equations

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.

2014-12-23

We show new solver strategies for the multigroup SPN equations for nuclear reactor analysis. By forming the complete matrix over space, moments, and energy a robust set of solution strategies may be applied. Moreover, power iteration, shifted power iteration, Rayleigh quotient iteration, Arnoldi's method, and a generalized Davidson method, each using algebraic and physics-based multigrid preconditioners, have been compared on C5G7 MOX test problem as well as an operational PWR model. These results show that the most ecient approach is the generalized Davidson method, that is 30-40 times faster than traditional power iteration and 6-10 times faster than Arnoldi's method.

Current and anticipated uses of thermal-hydraulic codes in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

Design and implementation of a simple nuclear power plant simulator

NASA Astrophysics Data System (ADS)

Miller, William H.

1983-02-01

A simple PWR nuclear power plant simulator has been designed and implemented on a minicomputer system. The system is intended for students use in understanding the power operation of a nuclear power plant. A PDP-11 minicomputer calculates reactor parameters in real time, uses a graphics terminal to display the results and a keyboard and joystick for control functions. Plant parameters calculated by the model include the core reactivity (based upon control rod positions, soluble boron concentration and reactivity feedback effects), the total core power, the axial core power distribution, the temperature and pressure in the primary and secondary coolant loops, etc.

Effects of the weld thermal cycle on the microstructure of alloy 690

NASA Astrophysics Data System (ADS)

Tuttle, James R.

Alloy 690 has been introduced as a material for use as the heat exchanger tubes in the steam generators (SGs) of pressurised water reactor (PWR) nuclear power plant. Its immediate predecessor, alloy 600, suffered from a number of degradation modes and another alternative, alloy 800, has also had in-service problems. In laboratory tests, alloy 690 in both mill annealed (MA) and special thermally treated (STT) condition has shown a high degree of resistance to degradation in simulated PWR primary side environments and other test media.Limited research has previously been undertaken to investigate the effects of welding on alloy 690, when the material is used in SG applications. It was deemed important to increase knowledge in this area since fabrication of PWR SGs involves gas tungsten arc welding (GTAW) of the heat exchanger tubes to a clad tubeplate. For this research investigation welded samples of alloy 690 have been produced in the laboratory using a range of thermal cycles based around recommended weld parameters for SG fabrication. These samples have been compared with archive welds from PWR SG manufacturers. A number of welds incorporating alloy 600 and a number using alloy 800 tubing material have also been fabricated in the laboratory for comparative purposes. Two experimental melts have been produced to study the effects of Nb substitution for Ti in alloy 690 type materials.Welded and unwelded specimens have been studied, analysed and tested using a variety of methods and techniques. A method of metallographic sample preparation for transmission electron microscope (TEM) thin foil specimens has been developed and documented which ensures foil perforation in a specific region. The effects of Nb substitution for Ti have been discussed. Chemical balances and microstructures in the fusion zone of welds manufactured from alloy 690 tubing incorporating alloy 82 weld consumable have been shown to be non-ideal. Within the heat affected zone (HAZ) of both laboratory produced and archive welds the microstructures have been identified as detrimentally altered from the STT condition original tubing material(s). A number of conclusions have been drawn and recommendations have been made for future work.

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code used for VVER in this research is Moby-Dick, which was developed to analyze the VVER by SKODA Inc. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1.

Corrosion performance of alternative steam generator materials and designs. Volume 2. Posttest examination of a seawater-faulted alternative materials model steam generator. Final report. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krupowicz, J.J.; Scott, D.B.; Fink, G.C.

Corrosion results obtained from the post-test non-destructive and destructive examinations of an alternative materials model steam generator are described in this final report. The model operated under representative thermal and hydraulic and accelerated (high seawater contaminant concentration) steam generator secondary water chemistry conditions. Total exposure consisted of 114 steaming days under all volatile treatment (AVT) chemistry conditions followed by 282 fault steaming days at a 30 ppM chloride concentration in the secondary bulk water. Various support plate and lattice strip support designs incorporated Types 347, 405, 409 and SCR-3 stainless steels; Alloys 600 and 690; and carbon steel. Heat transfermore » tube materials included Alloy 600 in various heat treated conditions, Alloy 690, and Alloy 800. All tubing materials in this test exhibited moderate pitting, primarily in the sludge pile region above the tubesheet.« less

Corrosion performance of alternative steam generator materials and designs. Volume 3. Posttest examination of a freshwater-faulted alternative materials model steam generator. Final report. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krupowicz, J.J.; Scott, D.B.; Rentler, R.M.

Corrosion results obtained from the post-test non-destructive and destructive examinations of an alternative materials model steam generator are described in this final report. The model operated under representative thermal and hydraulic and accelerated (high fresh water contaminant concentration) steam generator secondary water chemistry conditions. Total exposure consisted of 114 steaming days under all volatile treatment (AVT) chemistry conditions followed by 358 fault steaming days at a 40 ppM sulfate concentration in the secondary bulk water. Various support plate and lattice strip support designs incorporated Types 347, 405, 409 and SCR-3 stainless steels; Alloys 600 and 690; and carbon steel. Heatmore » transfer tube materials included Alloy 600 in various heat treated conditions, Alloy 690, and Alloy 800. All tubing materials in this test exhibited significant general corrosion beneath thick surface deposits.« less

Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. L. Davis; D. L. Knudson; J. L. Rempe

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status ofmore » INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.« less

Automatic treatment of the variance estimation bias in TRIPOLI-4 criticality calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dumonteil, E.; Malvagi, F.

2012-07-01

The central limit (CLT) theorem States conditions under which the mean of a sufficiently large number of independent random variables, each with finite mean and variance, will be approximately normally distributed. The use of Monte Carlo transport codes, such as Tripoli4, relies on those conditions. While these are verified in protection applications (the cycles provide independent measurements of fluxes and related quantities), the hypothesis of independent estimates/cycles is broken in criticality mode. Indeed the power iteration technique used in this mode couples a generation to its progeny. Often, after what is called 'source convergence' this coupling almost disappears (the solutionmore » is closed to equilibrium) but for loosely coupled systems, such as for PWR or large nuclear cores, the equilibrium is never found, or at least may take time to reach, and the variance estimation such as allowed by the CLT is under-evaluated. In this paper we first propose, by the mean of two different methods, to evaluate the typical correlation length, as measured in cycles number, and then use this information to diagnose correlation problems and to provide an improved variance estimation. Those two methods are based on Fourier spectral decomposition and on the lag k autocorrelation calculation. A theoretical modeling of the autocorrelation function, based on Gauss-Markov stochastic processes, will also be presented. Tests will be performed with Tripoli4 on a PWR pin cell. (authors)« less

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

Densities ............................................................................................................ 21 2.3 Fuel Mass (Core Total...70 7.1 Geometry, Material Density, and Mass Summary for All Cores...21 Table 3: Fuel Rod Masses for Different Clads

Self-referenced directional enhanced Raman scattering using plasmon waveguide resonance for surface and bulk sensing

NASA Astrophysics Data System (ADS)

Wan, Xiu-mei; Gao, Ran; Lu, Dan-feng; Qi, Zhi-mei

2018-01-01

Surface plasmon-coupled emission has been widely used in fluorescence imaging, biochemical sensing, and enhanced Raman spectroscopy. A self-referenced directional enhanced Raman scattering for simultaneous detection of surface and bulk effects by using plasmon waveguide resonance (PWR) based surface plasmon-coupled emission has been proposed and experimentally demonstrated. Raman scattering was captured on the prism side in Kretschmann-surface plasmon-coupled emission. The distinct penetration depths (δ) of the evanescent field for the transverse electric (TE) and transverse magnetic (TM) modes result in different detected distances of the Raman signal. The experimental results demonstrate that the self-referenced directional enhanced Raman scattering of the TE and TM modes based on the PWR can detect and distinguish the surface and bulk effects simultaneously, which appears to have potential applications in researches of chemistry, medicine, and biology.

IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilles Youinou; Andrea Alfonsi

2012-03-01

This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis,more » the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.« less

The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J.; Marshall, William B. J.; Ilas, Germina

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidancemore » in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.« less

Development of cement solidification process for sodium borate waste generated from PWR plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirofumi Okabe; Tatsuaki Sato; Yuichi Shoji

2013-07-01

A cement solidification process for treating sodium borate waste produced in pressurized water reactor (PWR) plants was studied. To obtain high volume reduction and high mechanical strength of the waste, simulated concentrated borate liquid waste with a sodium / boron (Na/B) mole ratio of 0.27 was dehydrated and powdered by using a wiped film evaporator. To investigate the effect of the Na/B mole ratio on the solidification process, a sodium tetraborate decahydrate reagent with a Na/B mole ratio of 0.5 was also used. Ordinary portland cement (OPC) and some additives were used for the solidification. Solidified cement prepared from powderedmore » waste with a Na/B mole ratio 0.24 and having a high silica sand content (silica sand/cement>2) showed to improved uniaxial compressive strength. (authors)« less

Hydrothermal synthesis of Ni 2FeBO 5 in near-supercritical PWR coolant and possible effects of neutron-induced 10B fission in fuel crud

NASA Astrophysics Data System (ADS)

Sawicki, Jerzy A.

2011-08-01

The hydrothermal synthesis of a nickel-iron oxyborate, Ni 2FeBO 5, known as bonaccordite, was investigated at pressures and temperatures that might occur at the surface of high-power fuel rods in PWR cores and in supercritical water reactors, especially during localized departures from nucleate boiling and dry-outs. The tests were performed using aqueous mixtures of nickel and iron oxides with boric acid or boron oxide, and as a function of lithium hydroxide addition, temperature and time of heating. At subcritical temperatures nickel ferrite NiFe 2O 4 was always the primary reaction product. High yield of Ni 2FeBO 5 synthesis started near critical water temperature and was strongly promoted by additions of LiOH up to Li/Fe and Li/B molar ratios in a range 0.1-1. The synthesis of bonaccordite was also promoted by other alkalis such as NaOH and KOH. The bonaccordite particles were likely formed by dissolution and re-crystallization by means of an intermediate nickel ferrite phase. It is postulated that the formation of Ni 2FeBO 5 in deposits of borated nickel and iron oxides on PWR fuel cladding can be accelerated by lithium produced in thermal neutron capture 10B(n,α) 7Li reactions. The process may also be aided in the reactor core by kinetic energy of α-particles and 7Li ions dissipated in the crud layer.

Electron Microscopy Characterizations and Atom Probe Tomography of Intergranular Attack in Alloy 600 Exposed to PWR Primary Water

NASA Astrophysics Data System (ADS)

Olszta, Matthew J.; Schreiber, Daniel K.; Thomas, Larry E.; Bruemmer, Stephen M.

Detailed examinations of intergranular attack (IGA) in alloy 600 were performed after exposure to simulated PWR primary water at 325°C for 500 h. High-resolution analyses of IGA characteristics were conducted on specimens with either a 1 µm diamond or 1200-grit SiC surface finish using scanning electron microscopy, transmission electron microscopy and atom probe tomography techniques. The diamond-polish finish with very little preexisting subsurface damage revealed attack of high-energy grain boundaries that intersected the exposed surface to depths approaching 2 µm. In all cases, IGA from the surface is localized oxidation consisting of porous, nanocrystalline MO-structure and spinel particles along with regions of faceted wall oxidation. Surprisingly, this continuous IG oxidation transitions to discontinuous, discrete Cr-rich sulfide particles up to 50 nm in diameter. In the vicinity of the sulfides, the grain boundaries were severely Cr depleted (to <1 at%) and enriched in S. The 1200 grit SiC finish surface exhibited a preexisting highly strained recrystallized layer of elongated nanocrystalline matrix grains. Similar IG oxidation and leading sulfide particles were found, but the IGA depth was typically confined to the near-surface ( 400 nm) recrystallized region. Difference in IGA for the two surface finishes indicates that the formation of grain boundary sulfides occurs during the exposure to PWR primary water. The source of S remains unclear, however it is not present as sulfides in the bulk alloy nor is it segregated to bulk grain boundaries.

76 FR 66090 - Facility Operating License Amendment From Virginia Electric and Power Company, Surry Power...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-25

... operating pressures, leakage from primary water stress corrosion cracking below the proposed limited... discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR [Pressurized-Water Reactor...

The effects of parameter variation on MSET models of the Crystal River-3 feedwater flow system.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miron, A.

1998-04-01

In this paper we develop further the results reported in Reference 1 to include a systematic study of the effects of varying MSET models and model parameters for the Crystal River-3 (CR) feedwater flow system The study used archived CR process computer files from November 1-December 15, 1993 that were provided by Florida Power Corporation engineers Fairman Bockhorst and Brook Julias. The results support the conclusion that an optimal MSET model, properly trained and deriving its inputs in real-time from no more than 25 of the sensor signals normally provided to a PWR plant process computer, should be able tomore » reliably detect anomalous variations in the feedwater flow venturis of less than 0.1% and in the absence of a venturi sensor signal should be able to generate a virtual signal that will be within 0.1% of the correct value of the missing signal.« less

ENEL overall PWR plant models and neutronic integrated computing systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pedroni, G.; Pollachini, L.; Vimercati, G.

1987-01-01

To support the design activity of the Italian nuclear energy program for the construction of pressurized water reactors, the Italian Electricity Board (ENEL) needs to verify the design as a whole (that is, the nuclear steam supply system and balance of plant) both in steady-state operation and in transient. The ENEL has therefore developed two computer models to analyze both operational and incidental transients. The models, named STRIP and SFINCS, perform the analysis of the nuclear as well as the conventional part of the plant (the control system being properly taken into account). The STRIP model has been developed bymore » means of the French (Electricite de France) modular code SICLE, while SFINCS is based on the Italian (ENEL) modular code LEGO. STRIP validation was performed with respect to Fessenheim French power plant experimental data. Two significant transients were chosen: load step and total load rejection. SFINCS validation was performed with respect to Saint-Laurent French power plant experimental data and also by comparing the SFINCS-STRIP responses.« less

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2010-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10{sup 2-4}), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herer, C.; Souyri, A.; Garnier, J.

1995-09-01

Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to themore » annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.« less

International Space Station United States Orbital Segment Oxygen Generation System On-Orbit Operational Experience

NASA Technical Reports Server (NTRS)

Erickson, Robert J.; Howe, John, Jr.; Kulp, Galen W.; VanKeuren, Steven P.

2008-01-01

The International Space Station (ISS) United States Orbital Segment (USOS) Oxygen Generation System (OGS) was originally intended to be installed in ISS Node 3. The OGS rack delivery was accelerated, and it was launched to ISS in July of 2006 and installed in the US Laboratory Module. Various modification kits were installed to provide its interfaces, and the OGS was first activated in July of 2007 for 15 hours, In October of 2007 it was again activated for 76 hours with varied production rates and day/night cycling. Operational time in each instance was limited by the quantity of feedwater in a Payload Water Reservoir (PWR) bag. Feedwater will be provided by PWR bag until the USOS Water Recovery System (WRS) is delivered to SS in fall of 2008. This paper will discuss operating experience and characteristics of the OGS, as well as operational issues and their resolution.

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

PWR steam generator chemical cleaning, Phase I. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rothstein, S.

1978-07-01

United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the searchmore » sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.« less

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1979-10-01

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminary design for a high-burnup fuel rod, physics analyses of extended-burnup fuel cycles, studies of the physics characteristics of changes in fuel assembly metal-to-water ratios, and developmentmore » of a design concept for post-irradiation examination equipment to be utilized in examining high-burnup lead-test assemblies.« less

Progress in understanding fission-product behaviour in coated uranium-dioxide fuel particles

NASA Astrophysics Data System (ADS)

Barrachin, M.; Dubourg, R.; Kissane, M. P.; Ozrin, V.

2009-03-01

Supported by results of calculations performed with two analytical tools (MFPR, which takes account of physical and chemical mechanisms in calculating the chemical forms and physical locations of fission products in UO2, and MEPHISTA, a thermodynamic database), this paper presents an investigation of some important aspects of the fuel microstructure and chemical evolutions of irradiated TRISO particles. The following main conclusions can be identified with respect to irradiated TRISO fuel: first, the relatively low oxygen potential within the fuel particles with respect to PWR fuel leads to chemical speciation that is not typical of PWR fuels, e.g., the relatively volatile behaviour of barium; secondly, the safety-critical fission-product caesium is released from the urania kernel but the buffer and pyrolytic-carbon coatings could form an important chemical barrier to further migration (i.e., formation of carbides). Finally, significant releases of fission gases from the urania kernel are expected even in nominal conditions.

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

NASA Astrophysics Data System (ADS)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Effective dominance of resistance of Spodoptera frugiperda to Bt maize and cotton varieties: implications for resistance management

PubMed Central

Horikoshi, Renato J.; Bernardi, Daniel; Bernardi, Oderlei; Malaquias, José B.; Okuma, Daniela M.; Miraldo, Leonardo L.; Amaral, Fernando S. de A. e; Omoto, Celso

2016-01-01

The resistance of fall armyworm (FAW), Spodoptera frugiperda, has been characterized to some Cry and Vip3A proteins of Bacillus thuringiensis (Bt) expressed in transgenic maize in Brazil. Here we evaluated the effective dominance of resistance based on the survival of neonates from selected Bt-resistant, heterozygous, and susceptible (Sus) strains of FAW on different Bt maize and cotton varieties. High survival of strains resistant to the Cry1F (HX-R), Cry1A.105/Cry2Ab (VT-R) and Cry1A.105/Cry2Ab/Cry1F (PW-R) proteins was detected on Herculex, YieldGard VT PRO and PowerCore maize. Our Vip3A-resistant strain (Vip-R) exhibited high survival on Herculex, Agrisure Viptera and Agrisure Viptera 3 maize. However, the heterozygous from HX-R × Sus, VT-R × Sus, PW-R × Sus and Vip-R × Sus had complete mortality on YieldGard VT PRO, PowerCore, Agrisure Viptera, and Agrisure Viptera 3, whereas the HX-R × Sus and Vip-R × Sus strains survived on Herculex maize. On Bt cotton, the HX-R, VT-R and PW-R strains exhibited high survival on Bollgard II. All resistant strains survived on WideStrike, but only PW-R and Vip-R × Sus survived on TwinLink. Our study provides useful data to aid in the understanding of the effectiveness of the refuge strategy for Insect Resistance Management of Bt plants. PMID:27721425

EMERALD REV.1. PWR Accident Activity Release

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1975-10-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« less

Effective dominance of resistance of Spodoptera frugiperda to Bt maize and cotton varieties: implications for resistance management

NASA Astrophysics Data System (ADS)

Horikoshi, Renato J.; Bernardi, Daniel; Bernardi, Oderlei; Malaquias, José B.; Okuma, Daniela M.; Miraldo, Leonardo L.; Amaral, Fernando S. De A. E.; Omoto, Celso

2016-10-01

The resistance of fall armyworm (FAW), Spodoptera frugiperda, has been characterized to some Cry and Vip3A proteins of Bacillus thuringiensis (Bt) expressed in transgenic maize in Brazil. Here we evaluated the effective dominance of resistance based on the survival of neonates from selected Bt-resistant, heterozygous, and susceptible (Sus) strains of FAW on different Bt maize and cotton varieties. High survival of strains resistant to the Cry1F (HX-R), Cry1A.105/Cry2Ab (VT-R) and Cry1A.105/Cry2Ab/Cry1F (PW-R) proteins was detected on Herculex, YieldGard VT PRO and PowerCore maize. Our Vip3A-resistant strain (Vip-R) exhibited high survival on Herculex, Agrisure Viptera and Agrisure Viptera 3 maize. However, the heterozygous from HX-R × Sus, VT-R × Sus, PW-R × Sus and Vip-R × Sus had complete mortality on YieldGard VT PRO, PowerCore, Agrisure Viptera, and Agrisure Viptera 3, whereas the HX-R × Sus and Vip-R × Sus strains survived on Herculex maize. On Bt cotton, the HX-R, VT-R and PW-R strains exhibited high survival on Bollgard II. All resistant strains survived on WideStrike, but only PW-R and Vip-R × Sus survived on TwinLink. Our study provides useful data to aid in the understanding of the effectiveness of the refuge strategy for Insect Resistance Management of Bt plants.

Hopkins during ITCS PWR Retrieval

NASA Image and Video Library

2014-01-31

ISS038-E-040140 (31 Jan. 2014) --- NASA astronaut Mike Hopkins, Expedition 38 flight engineer, uses the Fluid Servicing System (FSS) to refill Internal Thermal Control System (ITCS) loops with fresh coolant in the Destiny laboratory of the International Space Station.

Hopkins during ITCS PWR Retrieval

NASA Image and Video Library

2014-01-31

ISS038-E-040139 (31 Jan. 2014) --- NASA astronaut Mike Hopkins, Expedition 38 flight engineer, uses the Fluid Servicing System (FSS) to refill Internal Thermal Control System (ITCS) loops with fresh coolant in the Destiny laboratory of the International Space Station.

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

2017-01-12

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

On the Geological History of Venus

NASA Astrophysics Data System (ADS)

Basilevsky, A. T.; Head, J. W.

2008-09-01

mostly based on the analysis of data acquired by the Magellan mission: SAR images with 100-200 m resolution and the maps of topography, surface radar reflectivity, emissivity, roughness and gravity anomalies [1]. After initial analysis of the data summarized in [2, 3] several groups of researchers continued to study the geology and geophysics of the planet, resulting in numerous publications, some of which are referenced below. Very important for the studies emphasizing the geologic history of Venus was, and still is, a program of 1:5,000,000 geologic mapping coordinated by the US Geological Survey [4]. A recent summary of these studies can be found in [5]. Observations and analysis: All researchers in this study area analyze the same data sets and follow the same guidelines [4, 6] so geologic units identified by them and their time sequences are generally similar, although different researchers may name the same units differently and may interpret differently some details of local time sequences. Figure 1 shows a time sequence of geologic units suggested by [7, 8]: materials of tessera terrain (tt), densely fractured plains (pdf), fractured and ridged plains (pfr), shield plains (psh), plains with wrinkle ridges (pwr), lobate (pl) and smooth (ps) plains as well as materials of radar-dark craterassociated parabolas (cdp). These are material units. In addition, some researchers identify and map structural units. In Figure 1 examples of these are fracture belts (fb) and rifted terrain (rt). synchronous on a global scale. The first option can be visualized with Figure 1, suggesting that it is applicable for Venus globally. This option was suggested by Basilevsky and Head [e.g., 7, 8] as well as by Ivanov and Head [e.g., 9]. The second option, first clearly formulated by [10], can be visualized by the upper part of Figure 2 showing the situation in three different hypothetical geologic provinces on Venus. In these provinces the unit time sequences are the same: tt => pdf => pfr/RB => pwr, but morphologically similar units, for example, units pwr, are not synchronous between them: unit pwr in province 1 is generally synchronous with unit tt in province 2 and with unit pfr/RB in province 3. As it was mentioned in [7, 8], if geologic analysis and mapping are being done within spatially separated geologic provinces, the synchronous vs. not synchronous alternative cannot be resolved. But if the geologic analysis and mapping are done within large areas, which include several geologic provinces with nonsynchronous units sequences, then at the boundaries of the geologic provinces one should see contradictions in the units' age relations. These contradictions are visualized in the lower part of Figure 2: Tessera massif at the boundary between geologic provinces 1 and 2, from the province 1 side, should be formed as a result of tessera-forming deformation of material unit pwr, while from the province 2 side it is embayed by the unit pwr. Similar contradictions are observed in relations between tessera and unit pfr/RB in provinces 2 and 3. We have mentioned in several publications [e.g. 11] that very large (more than half) regions of the planet have been mapped and such contradictions were not met by us and not reported by other researchers. So we stated that this favored the synchronous option, but that time we could not global geologic mapping of Venus has been recently completed [12] and such contradictions have not been met, we can say that this mapping has proved that morphologically similar units occupying similar positions in the local time sequences are globally synchronous. Of course, each of units considered had been formed not instantaneously, but within some period of time. So we refer to the general synchroneity and minor overlapping in absolute time of formation between stratigraphically neighboring units as certainly possible. This global mapping of [12] led to the identification of geologic units and their time sequence that is very similar to those identified by [7, 8]; this allows us to return to that model of regional and global stratigraphy of Venus (Figure 3). This figure is almost identical to Figure 22 in [8] and differs only in the estimate of absolute age of the boundary between the Fortunian and Guineverian periods (1.2T here vs. 1.4T in [8]). The question of the estimation of absolute ages of geologic units is difficult for Venus because the atmosphere is too massive to allow craters smaller than a few kilometers in diameter to be formed on its surface. As a result, the total number of impact craters on Venus is only about 1,000, and this makes it possible to estimate more or less reliably only the mean surface age of Venus, and less reliably the mean ages of several large globally observed geologic units such as pwr, tt or pl. Crater count techniques used for other planetary bodies, which permits absolute dating and time correlations of units occupying relatively small areas, can not be used in this way on Venus. The existing estimates of mean absolute ages of the larger Venusian geologic units, such as pwr, tt or pl [e.g., 13-15], were obtained by counting craters on areally separated outcrops of these units and normalizing sums of crater numbers by the total areas of the unit outcrops. Although the results of such an approach were consistent with stratigraphies based on geologic analysis [e.g., 13-15] this was keeping in mind the possibility that in different part of the planet absolute ages of the same units may be significantly different. Now with the completion of the global geologic mapping of Venus, such inconsistencies are excluded. Another problem in crater-count-based estimations of absolute ages of the geological formations on Venus is related to uncertainties of a number of parameters crucial for reliable modeling to transition from number of craters to millions and billions of years: e.g., 1) the meteoroid flux in the vicinity of the planet Venus, 2) the physics of passing of meteoroids through the dense Venus atmosphere, and 3) cratering under the high atmospheric pressure. As a result of these uncertainties, even statistically reliable estimates of the mean surface age of Venus are not very certain: ~750 m.y., but any values between 300 m.y. and 1 b.y. are considered possible [16]. This is why researchers using crater statistics to estimate absolute ages of individual geologic units use fractions or multiples of the mean surface age of Venus, instead of millions or billions of years, designating it as T [e.g., 14, 15]. The new estimates based on the global geologic mapping of [12] suggest that ages of selected units (in T with 2 σ error bars) are: tt, 1.09 ± 0.17; psh, 1.04 ± 0.18; pwr, 1.05 ± 0.12; pl, 0.54 ± 0.19; rt, 0.63 ± 0.26; confirming a new and more reliable basis for earlier estimates [13-15, 17-24]. This returns us to the conclusion made in [11]: …the earlier suite of units (from heavily deformed tesserae through slightly deformed regional plains) occurred during a time period an order of magnitude shorter than the subsequent period (from the end of emplacement of the wrinkle-ridge network until the present). These results imply high global rates of endogenic (volcanic) activity during the first era (comparable to that of mid-oceanic-ridge volcanism of Earth) and much lower global rates of endogenic activity (by two orders of magnitude) for the second period (page 1015, abstract). Conclusions: As it follows from the above consideration, tectonic and volcanic processes in the beginning of the morphologically recognizable part of the geologic history of Venus (since tessera time) were rather active and resurfaced the entire planet. But then, after about 10-20% of the total duration of this part of history they rather sharply occurred sporadically and in separate spots and zones and affected only 15-20% of the Venus. For better understanding the rates of tectonic and volcanic processes on Venus, knowledge of which is crucial for working out reliable geodynamic models of the evolution of this planet, we need to have isotopic dating for absolute ages of major geologic units. The most promising in this respect is a sample return mission to Venus aiming to return to Earth material of unit pwr [25]. References: [1] Saunders R.S. et al. (1992) JGR, 97, 13067- 13091. [2] JGR (1992) 97, E8, E10. [3] Venus II (1997) Univ. Arizona Press. 1362 p. [4] Tanaka K.L. (1994) USGS Open-File Report 94-438. [5] Basilevsky A.T. & McGill G.E. (2007) In: Exploring Venus as a Terrestrial Planet, Geophysical Monograph 176. American Geophysical Union, Washington, DC. 23-44. [6] Wilhelms D. (1990) in Planetary Mapping, NY, 208-260. [7] Basilevsky A.T. & Head J.W. (1998) JGR, 103, 8531-8544. [8] Basilevsky A.T. & Head J.W. (2000) PSS, 48, 75-111. [9] Ivanov & Head J.W. (2001) JGR, 106, 17515-17566. [10] Guest J.E. & Stofan E.E. [1999] Icarus, 139, 55-66. [11] Basilevsky A.T. & Head J.W. (2002a) Geology, 30, 1015-1018. [12] Ivanov M.A. (2008) LPSC XXXIX, abs. # 1017. [13] Ivanov M. A. & Basilevsky A.T. (1993) GRL, 20, 2579-2582. [14] Namiki, N. & Solomon S.C. (1994) Science, 265, 929-933. [15] Price, M. & Suppe J. (1994) Nature, 372, 756-759. [16] McKinnon W. et al. (1997) Venus II, Univ. Arizona Press, 969-1014. [17] Gilmore M.S. et al. (1997) JGR, 102, 13,357-13,368. [18] Collins G.C. (1999) JGR, 104, 24,121-24,139. [19] Basilevsky A.T. et al. (1999) GRL, 26, 2593-2596. [20] Pivchenkova E.V. & Kryuchkov V.P. (2001) Vernadsky- Brown Microsymposium 34, abs. MS057. [21] Basilevsky A.T. & Head J.W. (2002b) JGR, 107, doi: 10.1029/2000JE001471. [22] Basilevsky A.T. & Head J.W. (2002c) JGR, 107, 10.1029/2001JE001584, 2002. [23] McGill G.E. (2004) Icarus, 172, 603-612. [24] Basilevsky A.T. & Head J.W. (2006) JGR, 111, CiteID E03006. [25] Basilevsky A.T. et al. (2006) PSS, 55, 2097-2112.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.

The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue frommore » J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code.« less

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

EMERALD REVISION 1; PWR accident activity release. [IBM360,370; FORTRAN IV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fowler, T.B.; Tobias, M.L.; Fox, J.N.

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.IBM360,370; FORTRAN IV; OS/360,370 (IBM360,370); 520K bytes of memory are required..« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Probabilistic analysis on the failure of reactivity control for the PWR

NASA Astrophysics Data System (ADS)

Sony Tjahyani, D. T.; Deswandri; Sunaryo, G. R.

2018-02-01

The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria.

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure thatmore » critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that significantly exceeded QE limits for extended time periods for the low frequency STSBO sequence evaluated in this study. It is recognized that the core damage frequency (CDF) of the sequence evaluated in this scoping effort would be considerably lower if evaluations considered new FLEX equipment being installed by industry. Nevertheless, because of uncertainties in instrumentation response when exposed to conditions beyond QE limits and alternate challenges associated with different sequences that may impact sensor performance, it is recommended that additional evaluations of instrumentation performance be completed to provide confidence that operators have access to accurate, relevant, and timely information on the status of reactor systems for a broad range of challenges associated with risk important severe accident sequences.« less

78 FR 35960 - Minor Boundary Revision at Mojave National Preserve

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-14

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-MOJA-12321; PS.SMOJA0003] Minor Boundary Revision at Mojave National Preserve AGENCY: National Park Service, Interior. ACTION: Notification of... following locations: National Park Service, Land Resources Program Center, Pacific West Region, 333 Bush...

Grain boundary damage evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhai, Ziqing; Toloczko, Mychailo B.; Kruska, Karen

Long-term grain boundary (GB) damage evolution and stress corrosion crack initiation in alloy 690 are being investigated by constant load tensile testing in high-temperature, simulated PWR primary water. Six commercial alloy 690 heats are being tested in various cold work conditions loaded at their yield stress. This paper reviews the basic test approach and detailed characterizations performed on selected specimens after an exposure time of ~1 year. Intergranular crack nucleation was observed under constant stress in certain highly cold-worked (CW) alloy 690 heats and was found to be associated with the formation of GB cavities. Somewhat surprisingly, the heats mostmore » susceptible to cavity formation and crack nucleation were thermally treated materials with most uniform coverage of small GB carbides. Microstructure, % cold work and applied stress comparisons are made among the alloy 690 heats to better understand the factors influencing GB cavity formation and crack initiation.« less

Microstructural Effects on SCC Initiation PWR Primary Water Cold-Worked Alloy 600

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhai, Ziqing; Toloczko, Mychailo B.; Bruemmer, Stephen M.

SCC initiation behavior of one mill annealed alloy 600 plate heat was investigated in simulated PWR primary water under constant load at yield stress with in-situ direct current potential drop (DCPD) monitoring for crack initiation. Twelve specimens were tested at similar cold work levels among which three showed much shorter SCC initiation times (<400 hrs) than the others (>1200 hrs). Post-test examinations revealed that these three specimens all feature an inhomogeneous microstructure where the primary crack always nucleated along the boundary of large elongated grains protruding normally into the gauge. In contrast, such microstructure was either not observed or didmore » not extend deep enough into the gauge in the other specimens exhibiting ~3-6X longer initiation times. In order to better understand the role of this microstructural inhomogeneity in SCC initiation, high-resolution microscopy was performed to compare carbide morphology and strain distribution between the long grains and normal grains, and their potential effects on SCC initiation are discussed in this paper.« less

Reactor antineutrino detector iDREAM.

NASA Astrophysics Data System (ADS)

Gromov, M. B.; Lukyanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

2017-09-01

Industrial Detector for Reactor Antineutrino Monitoring (iDREAM) is a compact (≈ 3.5m 2) industrial electron antineutrino spectrometer. It is dedicated for remote monitoring of PWR reactor operational modes by neutrino method in real-time. Measurements of antineutrino flux from PWR allow to estimate a fuel mixture in active zone and to check the status of the reactor campaign for non-proliferation purposes. LAB-based gadolinium doped scintillator is exploited as a target. Multizone architecture of the detector with gamma-catcher surrounding fiducial volume and plastic muon veto above and below ensure high efficiency of IBD detection and background suppression. DAQ is based on Flash ADC with PSD discrimination algorithms while digital trigger is programmable and flexible due to FPGA. The prototype detector was started up in 2014. Preliminary works on registration Cerenkov radiation produced by cosmic muons were established with distilled water inside the detector in order to test electronic and slow control systems. Also in parallel a long-term measurements with different scintillator samples were conducted.

Regeneratively Cooled Liquid Oxygen/Methane Technology Development

NASA Technical Reports Server (NTRS)

Robinson, Joel W.; Greene, Christopher B.; Stout, Jeffrey

2012-01-01

The National Aeronautics & Space Administration (NASA) has identified Liquid Oxygen (LOX)/Liquid Methane (LCH4) as a potential propellant combination for future space vehicles based upon exploration studies. The technology is estimated to have higher performance and lower overall systems mass compared to existing hypergolic propulsion systems. NASA-Marshall Space Flight Center (MSFC) in concert with industry partner Pratt & Whitney Rocketdyne (PWR) utilized a Space Act Agreement to test an oxygen/methane engine system in the Summer of 2010. PWR provided a 5,500 lbf (24,465 N) LOX/LCH4 regenerative cycle engine to demonstrate advanced thrust chamber assembly hardware and to evaluate the performance characteristics of the system. The chamber designs offered alternatives to traditional regenerative engine designs with improvements in cost and/or performance. MSFC provided the test stand, consumables and test personnel. The hot fire testing explored the effective cooling of one of the thrust chamber designs along with determining the combustion efficiency with variations of pressure and mixture ratio. The paper will summarize the status of these efforts.

On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

NASA Astrophysics Data System (ADS)

Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

2016-12-01

Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

Effects of iron content in Ni-Cr-xFe alloys and immersion time on the oxide films formed in a simulated PWR water environment

NASA Astrophysics Data System (ADS)

Ru, Xiangkun; Lu, Zhanpeng; Chen, Junjie; Han, Guangdong; Zhang, Jinlong; Hu, Pengfei; Liang, Xue

2017-12-01

The iron content in Ni-Cr-xFe (x = 0-9 at.%) alloys strongly affected the properties of oxide films after 978 h of immersion in the simulated PWR primary water environment at 310 °C. Increasing the iron content in the alloys increased the amount of iron-bearing polyhedral spinel oxide particles in the outer oxide layer and increased the local oxidation penetrations into the alloy matrix from the chromium-rich inner oxide layer. The effects of iron content in the alloys on the oxide film properties after 500 h of immersion were less significant than those after 978 h. Iron content increased, and chromium content decreased, in the outer oxide layer with increasing iron content in the alloys. Increasing the immersion time facilitated the formation of the local oxidation penetrations along the matrix/film interface and the nickel-bearing spinel oxides in the outer oxide layer.

Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.

1982-12-01

During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency andmore » R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr; Park, Sangrok; Kim, Byong Sup

Since the first nuclear power was engaged in Korean electricity grid in 1978, intensive research and development has been focused on localization and standardization of large pressurized water reactors (PWRs) aiming at providing Korean peninsula and beyond with economical and safe power source. With increased priority placed on the safety since Chernobyl accident, Korean nuclear power R and D activity has been diversified into advanced PWR, small modular PWR and generation IV reactors. After the outbreak of Fukushima accident, inherently safe small modular reactor (SMR) receives growing interest in Korea and Europe. In this paper, we will describe recent statusmore » of evolving designs of SMR, their advantages and challenges. In particular, the conceptual design of lead-bismuth cooled SMR in Korea, URANUS with 40∼70 MWe is examined in detail. This paper will cover a framework of the program and a strategy for the successful deployment of small modular reactor how the goals would entail and the approach to collaboration with other entities.« less

Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions

NASA Astrophysics Data System (ADS)

Åberg Lindell, M.; Andersson, P.; Grape, S.; Håkansson, A.; Thulin, M.

2018-07-01

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles. One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties. From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively. An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft..., ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components.'' This draft LR-ISG revises...

77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-19

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft...-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components...

Proceedings: 2002 Workshop on Pressurized Water Reactor Elevated Feedwater Iron Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2002-11-01

Some pressurized water reactor (PWR) stations have experienced difficulty with maintaining feedwater (FW) iron concentrations below recommended concentration on a regular basis. A workshop held on September 17-18 in Dana Point, California, addressed the challenge of elevated feedwater iron transport in PWRs.

75 FR 15752 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-30

...)--191, Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public... Fuel Cycle Oversight Process Revisions (Public Meeting). (Contact: Michael Raddatz, 301-492-3108.) This..., Employee/Labor Relations and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e-mail at angela...

75 FR 18907 - Sunshine Federal Register Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-13

...)-191, Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public... Fuel Cycle Oversight Process Revisions (Public Meeting). (Contact: Michael Raddatz, 301-492-3108.) This... Bolduc, Chief, Employee/Labor Relations and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e...

Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog

The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to addressmore » this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.« less

Final Prep on SSME

NASA Image and Video Library

2005-10-25

Alvin Pittman Sr., lead electronics technician with Pratt & Whitney Rocketdyne, and Janine Cuevas, a mechanical technician with PWR, perform final preparations on the space shuttle main engine tested Oct. 25, 2005, at NASA's Stennis Space Center. It was the first main engine test since Hurricane Katrina hit the Gulf Coast on Aug. 29.

Final Prep on SSME

NASA Technical Reports Server (NTRS)

2005-01-01

Alvin Pittman Sr., lead electronics technician with Pratt & Whitney Rocketdyne, and Janine Cuevas, a mechanical technician with PWR, perform final preparations on the space shuttle main engine tested Oct. 25, 2005, at NASA's Stennis Space Center. It was the first main engine test since Hurricane Katrina hit the Gulf Coast on Aug. 29.

DEVELOPMENT OF WELDED SEAL FOR S3G REACTOR VESSEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rogers, J.W.

1958-01-01

The development program consisted of preliminary design, welding accessibility and feasibility, pressure and displacement cycling, theoretical analysis and life computation, photoelastic analysis, and comparison of PWR straight sample cycling. Design ''C'' of the three primary designs considered proved more satisfactory from a fatigue life standpoint. (W.D. M.)

76 FR 40937 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-12

... Generator Water Level High-High'' instrument setpoint and associated allowable value. The proposed change is... [Pressurized-Water Reactor] PWR Operability Requirements and Actions for RCS Leakage Instrumentation''. Basis... monitor is the containment atmospheric gaseous radiation monitor. The monitoring of RCS leakage is not a...

76 FR 21917 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-19

... SGTR accident. At normal operating pressures, leakage from primary water stress corrosion cracking... PWR [pressurized- water reactor] Operability Requirements and Actions for RCS Leakage Instrumentation... water inventory can be obtained. Therefore, it is concluded that the proposed changes do not involve a...

76 FR 1644 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-11

... tubesheet in that region. At normal operating pressures, leakage from primary water stress corrosion... cause failure. The EDG reliability will thereby be potentially increased by reducing the stresses on the..., ``Bases for Plugging Degraded PWR [pressurized-water reactor] Steam Generator Tubes,'' margins against...

Spelling well Despite Developmental Language Disorder: What Makes It Possible?

ERIC Educational Resources Information Center

Rakhlin, Natalia; Cardoso-Martins, Cláudia; Kornilov, Sergey A.; Grigorenko, Elena L.

2013-01-01

The goal of the study was to investigate the overlap between developmental language disorder (DLD) and developmental dyslexia, identified through spelling difficulties (SD), in Russian-speaking children. In particular, we studied the role of phoneme awareness (PA), rapid automatized naming (RAN), pseudoword repetition (PWR), morphological (MA),…

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

78 FR 44596 - Minor Boundary Revision at Yosemite National Park

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-24

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-YOSE-13178; PS.SPWLA0028.00.1] Minor Boundary Revision at Yosemite National Park AGENCY: National Park Service, Interior. ACTION: Notification of Boundary Revision. SUMMARY: The boundary of Yosemite National Park is modified to include 80 acres...

78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0268] Review of Safety Analysis Reports for Nuclear Power... Analysis Reports for Nuclear Power Plants: LWR Edition.'' The new subsection is the Standard Review Plan... Nuclear Power Plants: Integral Pressurized Water Reactor (iPWR) Edition.'' DATES: Comments must be filed...

77 FR 63343 - Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-16

... PWR [Pressurized-Water Reactor] Steam Generator Tubes'' (Reference 32) and [Nuclear Energy Institute... maintains the required structural margins of the SG tubes for both normal and accident conditions. Nuclear Energy Institute 97-06, ``Steam Generator Program Guidelines'' (Reference 8), and NRC Regulatory Guide 1...

Counterfeit Parts Prevention Strategy Guide Product Overview

DTIC Science & Technology

2014-05-08

pwr.utc.com Mark King Micopac markking@micropac.com Andrew King Boeing andrew.m.king@boeing.com Byron Knight NRO knightby@nro.mil Hans Koenigsmann SpaceX ...Marvin VanderWeg SpaceX marvin.vanderwag@spacex.com Gerrit VanOmmering SSL gerrit.vanommering@sslmda.com Michael Verzuh Ball mverzuh@ball.com John Vilja

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

NASA Astrophysics Data System (ADS)

Seguin, Nicolas

2014-05-01

In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical model, this numerical scheme is also efficient in terms of CPU time. Eventually, simpler models can locally replace the more complex model in order to simplify the overall computation, using some appropriate local error indicators developed in [5], without reducing the accuracy. References 1. Ishii, M., Hibiki, T., Thermo-fluid dynamics of two-phase flow, Springer, New-York, 2006. 2. Gallouët, T. and Hérard, J.-M., Seguin, N., Numerical modeling of two-phase flows using the two-fluid two-pressure approach, Math. Models Methods Appl. Sci., Vol. 14, 2004. 3. Seguin, N., Étude d'équations aux dérivées partielles hyperboliques en mécanique des fluides, Habilitation à diriger des recherches, UPMC-Paris 6, 2011. 4. Coquel, F., Hérard, J-M., Saleh, K., Seguin, N., A Robust Entropy-Satisfying Finite Volume Scheme for the Isentropic Baer-Nunziato Model, ESAIM: Mathematical Modelling and Numerical Analysis, Vol. 48, 2013. 5. Mathis, H., Cancès, C., Godlewski, E., Seguin, N., Dynamic model adaptation for multiscale simulation of hyperbolic systems with relaxation, preprint, 2013.

Job analysis of maintenance-mechanic position for the nuclear power plant maintenance personnel reliability model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siegel, A.I.; Bartter, W.D.; Kopstein, F.F.

1982-06-01

The task list method of job survey was used. In collaboration with BWR and PWR personnel, a list of 107 tasks performed by maintenance mechanics was developed, grouped into: remove and install, test and repair, inspect and perform preventive maintenance, miscellaneous, communication, and report preparation. For each listed task, the questionnaire form inquired into: frequency of performance, task completion time, safety consequences of improper performance, and the amount of training required to perform the task proficiently. Scaled information was requested about seven abilities: (1) visual speed, accuracy, and recognition; (2) gross motor coordination; (3) fine manual dexterity; (4) strength andmore » stamina; (5) cognition; (6) memory; and (7) problem solving required for function completion. Survey forms were distributed to 27 nuclear power plants. Thirty-one maintenance mechanics representing 17 plants returned the completed forms. Frequency of performing tasks was bimodally distributed: (1) between once a year and once every six months, and (2) about once a week. More than half of the tasks have potential risk consequences if improperly performed. The five tasks with the greatest risk implications in the case of inadequate performance were: (1) remove and install reactor and dry-well heads, (2) test and repair reactor system components, (3) remove and install pressurizer mechanical relief valves, (4) test and repair pressurizer relief valves, (5) remove and install core spray pumps, seals, and valves. Hierarchically, the public risk associated with the various functions was: (1) remove and install, (2) test and repair, (3) preventive maintenance, (4) miscellaneous tasks, (5) communication, and (6) report preparation.« less

RELAP5 Analyses of OECD/NEA ROSA-2 Project Experiments on Intermediate-Break LOCAs at Hot Leg or Cold Leg

NASA Astrophysics Data System (ADS)

Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo

Experiments simulating PWR intermediate-break loss-of-coolant accidents (IBLOCAs) with 17% break at hot leg or cold leg were conducted in OECD/NEA ROSA-2 Project using the Large Scale Test Facility (LSTF). In the hot leg IBLOCA test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing (LSC) induced by steam condensation on accumulator coolant injected into cold leg. Water remained on upper core plate in upper plenum due to counter-current flow limiting (CCFL) because of significant upward steam flow from the core. In the cold leg IBLOCA test, core dryout took place due to rapid liquid level drop in the core before LSC. Liquid was accumulated in upper plenum, steam generator (SG) U-tube upflow-side and SG inlet plenum before the LSC due to CCFL by high velocity vapor flow, causing enhanced decrease in the core liquid level. The RELAP5/MOD3.2.1.2 post-test analyses of the two LSTF experiments were performed employing critical flow model in the code with a discharge coefficient of 1.0. In the hot leg IBLOCA case, cladding surface temperature of simulated fuel rods was underpredicted due to overprediction of core liquid level after the core uncovery. In the cold leg IBLOCA case, the cladding surface temperature was underpredicted too due to later core uncovery than in the experiment. These may suggest that the code has remaining problems in proper prediction of primary coolant distribution.

On-line measurement of gaseous iodine species during a PWR severe accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haykal, I.; Doizi, D.; Perrin, A.

A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still notmore » clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)« less

Simulation of a main steam line break with steam generator tube rupture using trace

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallardo, S.; Querol, A.; Verdu, G.

A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation ofmore » the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)« less

Recent improvements of reactor physics codes in MHI

NASA Astrophysics Data System (ADS)

Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

2015-12-01

This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

Sensitivity Analysis of OECD Benchmark Tests in BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swiler, Laura Painton; Gamble, Kyle; Schmidt, Rodney C.

2015-09-01

This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on sensitivity analysis of a fuels performance benchmark problem. The benchmark problem was defined by the Uncertainty Analysis in Modeling working group of the Nuclear Science Committee, part of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD ). The benchmark problem involv ed steady - state behavior of a fuel pin in a Pressurized Water Reactor (PWR). The problem was created in the BISON Fuels Performance code. Dakota was used to generate and analyze 300 samples of 17 input parameters defining coremore » boundary conditions, manuf acturing tolerances , and fuel properties. There were 24 responses of interest, including fuel centerline temperatures at a variety of locations and burnup levels, fission gas released, axial elongation of the fuel pin, etc. Pearson and Spearman correlatio n coefficients and Sobol' variance - based indices were used to perform the sensitivity analysis. This report summarizes the process and presents results from this study.« less

Investigation of on-line chelant addition to PWR steam generators. Annual report, 1981

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tvedt, T.J.; Wallace, S.L.; Griffin, F. Jr.

1982-11-01

The thermostability of both ethylenediaminetetraacetic acid (EDTA) and hydroxyethylethylenediamininetriacetic acid (HEDTA) metal chelates in all volatile treatment water chemistry (AVT) was shown to be greater than or equal to thermostability of EDTA metal chelates in phosphate-sulfite water chemistry. HEDTA metal chelates were shown to have a much greater stability than EDTA metal chelates. Using samples taken from the EDTA metal chelate thermostability studies and samples from Commonwealth Research Corporation (CRC) model steam generators (MSG), EDTA decomposition products were determined. Active metal surfaces were shown to become passivated when exposed to EDTA and HEDTA concentrations as high as 0.1% w/w inmore » AVT. Trace amounts of iron in the water were found to increase the rate of passivation. Material balance and visual inspection data from CRC model steam generators showed that metal is being transported through and cleaning from the MSG's. EDTA metal chelates were removed from chelate solutions by passing the solutions over strong anion exchange resins.« less

Recent improvements of reactor physics codes in MHI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki

2015-12-31

This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipatedmore » transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.« less

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

76 FR 79708 - Draft Environmental Impact Statement/General Management Plan, Golden Gate National Recreation...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-22

...). The Plan/DEIS evaluates four alternatives for updating the current approach to management in Golden.... In recognition of the complexity of the proposed plan alternatives, and with deference to interest... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-1108-8862; 2031-A038-409] Draft...

78 FR 33120 - Final Interim Staff Guidance LR-ISG-2011-04; Updated Aging Management Criteria for Reactor Vessel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

..., ``Generic Aging Lessons Learned Report'' (GALL Report), for the aging management of Pressurized Water... communicate insights and lessons learned and to address emergent issues not covered in license renewal... ensure that PWR license renewal applicants will adequately address age-related degradation and aging...

LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

40 CFR 59.505 - How do I demonstrate compliance with the reactivity limits?

Code of Federal Regulations, 2012 CFR

2012-07-01

... (CONTINUED) AIR PROGRAMS (CONTINUED) NATIONAL VOLATILE ORGANIC COMPOUND EMISSION STANDARDS FOR CONSUMER AND COMMERCIAL PRODUCTS National Volatile Organic Compound Emission Standards for Aerosol Coatings § 59.505 How..., 2B, or 2C. WFi = weight fraction of component i in the product, (2) Calculate the PWR of each product...

40 CFR 59.505 - How do I demonstrate compliance with the reactivity limits?

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) AIR PROGRAMS (CONTINUED) NATIONAL VOLATILE ORGANIC COMPOUND EMISSION STANDARDS FOR CONSUMER AND COMMERCIAL PRODUCTS National Volatile Organic Compound Emission Standards for Aerosol Coatings § 59.505 How..., 2B, or 2C. WFi = weight fraction of component i in the product, (2) Calculate the PWR of each product...

40 CFR 59.505 - How do I demonstrate compliance with the reactivity limits?

Code of Federal Regulations, 2014 CFR

2014-07-01

... (CONTINUED) AIR PROGRAMS (CONTINUED) NATIONAL VOLATILE ORGANIC COMPOUND EMISSION STANDARDS FOR CONSUMER AND COMMERCIAL PRODUCTS National Volatile Organic Compound Emission Standards for Aerosol Coatings § 59.505 How..., 2B, or 2C. WFi = weight fraction of component i in the product, (2) Calculate the PWR of each product...

75 FR 13800 - Sunshine Federal Register Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-23

..., Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public Meeting..., 2010. Week of April 26, 2010--Tentative Thursday, April 29, 2010 9:30 a.m. Briefing on the Fuel Cycle... and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e-mail at [email protected

77 FR 65906 - Minor Boundary Revision at Minidoka National Historic Site

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-31

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-MIIN-11234; 9360-726] Minor Boundary Revision at Minidoka National Historic Site AGENCY: National Park Service, Interior. ACTION: Notification of boundary revision. SUMMARY: Notice is hereby given that, pursuant to 16 U.S.C. 460l- 9(c)(1)(ii...

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

76 FR 24514 - Honouliuli Special Resource Study, Honolulu, Maui, Hawaii, and Kauai Counties, HI

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-02

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-0308-6923;9082-HONO-420] Honouliuli.... Background: As authorized by the Department of the Interior, Environment, and Related Agencies Appropriations... State of Hawaii with respect to (1) Their significance as components of World War II; (2) significance...

76 FR 22917 - Dog Management Plan/Draft Environmental Impact Statement, Golden Gate National Recreation Area...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-25

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO--0315-696; 8145-8B90-SZM] Dog... Impact Statement/Dog Management Plan, Golden Gate National Recreation Area. SUMMARY: The National Park Service has prepared a Draft Dog Management Plan and Environmental Impact Statement (Plan/DEIS). The Plan...

Amorphous and Nanocrystalline High Temperature Magnetic Material for PWR

DTIC Science & Technology

2006-03-01

FOR PUBLICATION. //Signature// //Signature// ______________________________________ __________________________________ JOHN C ...times that of conventional ferrites at room temperature); 2) Frequency: 200 kHz to 1 MHz; 3) Temperature: 200 ° C and above. The goals of the DUST...NAME OF RESPONSIBLE PERSON (Monitor) a. REPORT Unclassified b. ABSTRACT Unclassified c . THIS PAGE Unclassified 17. LIMITATION OF ABSTRACT

Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold Rolled Alloy 690 in PWR Primary Water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bruemmer, Stephen M.; Olszta, Matthew J.; Toloczko, Mychailo B.

2012-10-01

Grain boundary microstructures and microchemistries are examined in cold-rolled alloy 690 tubing and plate materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in PWR primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. Formore » the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results suggest that the localized grain boundary strains and stresses produced during cold rolling promote IGSCC susceptibility and not the cracked carbides and voids.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Overman, Nicole R.; Toloczko, Mychailo B.; Olszta, Matthew J.

High chromium, nickel-base Alloy 690 exhibits an increased resistance to stress corrosion cracking (SCC) in pressurized water reactor (PWR) primary water environments over lower chromium alloy 600. As a result, Alloy 690 has been used to replace Alloy 600 for steam generator tubing, reactor pressure vessel nozzles and other pressure boundary components. However, recent laboratory crack-growth testing has revealed that heavily cold-worked Alloy 690 materials can become susceptible to SCC. To evaluate reasons for this increased SCC susceptibility, detailed characterizations have been performed on as-received and cold-worked Alloy 690 materials using electron backscatter diffraction (EBSD) and Vickers hardness measurements. Examinationsmore » were performed on cross sections of compact tension specimens that were used for SCC crack growth rate testing in simulated PWR primary water. Hardness and the EBSD integrated misorientation density could both be related to the degree of cold work for materials of similar grain size. However, a microstructural dependence was observed for strain correlations using EBSD and hardness which should be considered if this technique is to be used for gaining insight on SCC growth rates« less

Probability of in-vessel steam explosion-induced containment failure for a KWU PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Esmaili, H.; Khatib-Rahbar, M.; Zuchuat, O.

During postulated core meltdown accidents in light water reactors, there is a likelihood for an in-vessel steam explosion when the melt contacts the coolant in the lower plenum. The objective of the work described in this paper is to determine the conditional probability of in-vessel steam explosion-induced containment failure for a Kraftwerk Union (KWU) pressurized water reactor (PWR). The energetics of the explosion depends on the mass of the molten fuel that mixes with the coolant and participates in the explosion and on the conversion of fuel thermal energy into mechanical work. The work can result in the generation ofmore » dynamic pressures that affect the lower head (and possibly lead to its failure), and it can cause acceleration of a slug (fuel and coolant material) upward that can affect the upper internal structures and vessel head and ultimately cause the failure of the upper head. If the upper head missile has sufficient energy, it can reach the containment shell and penetrate it. The analysis, must therefore, take into account all possible dissipation mechanisms.« less

From Paper to Production to Test: An Update on NASA's J-2X Engine for Exploration

NASA Technical Reports Server (NTRS)

Kynard, Michael

2011-01-01

The NASA/industry team responsible for developing the J-2X upper stage engine for the Space Launch System (SLS) Program has made significant progress toward moving beyond the design phase and into production, assembly, and test of development hardware. The J-2X engine exemplifies the SLS Program goal of using proven technology and experience from more than 50 years of United States spaceflight experience combined with modern manufacturing processes and approaches. It will power the second stage of the fully evolved SLS Program launch vehicle that will enable a return to human exploration of space beyond low earth orbit. Pratt & Whitney Rocketdyne (PWR) is under contract to develop and produce the engine, leveraging its flight-proven LH2/LOX, gas generator cycle J-2 and RS-68 engine capabilities, recent experience with the X-33 aerospike XRS-2200 engine, and development knowledge of the J-2S tap-off cycle engine. The J- 2X employs a gas generator operating cycle designed to produce 294,000 pounds of vacuum thrust in primary operating mode with its full nozzle extension. With a truncated nozzle extension suitable to support engine clustering on the stage, the nominal vacuum thrust level in primary mode is 285,000 pounds. It also has a secondary mode, during which it operates at 80 percent thrust by altering its mixture ratio. The J-2X development philosophy is based on proven hardware, an aggressive development schedule, and early risk reduction. NASA Marshall Space Flight Center (MSFC) and PWR began development of the J-2X in June 2006. The government/industry team of more than 600 people within NASA and PWR successfully completed the Critical Design Review (CDR) in November 2008, following extensive risk mitigation testing. Assembly of the first development engine was completed in May 2011 and the first engine test was conducted at the NASA Stennis Space Center (SSC), test stand A2, on 14 July 2011. Testing of the first development engine will continue through the autumn of 2011, be paused for test stand modifications to the passive diffuser, and then restart in the spring of 2012. This testing will be followed by specialized powerpack testing intended to examine the design and operating margins of the engine turbomachinery. The development plan beyond this point leads through more system-level, engine testing of several samples, analytical model validation activities, functional and performance verification, and then ultimate certification to support human spaceflight. This paper will discuss the J-2X development background, provide top-level information on design and development planning, and will explore some of the development challenges and mitigation activities pursued to date.

Characterization of carbon-14 generated by the nuclear power industry. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eabry, S.; Vance, J.N.; Cline, J.E.

1995-11-01

This report describes an evaluation of C-14 production rates in light-water reactors (LWRs) and characterization of its chemical speciation and environmental behavior. The study estimated the total production rate of the nuclide in operating PWRs and BWRs along with the assessment of the C-14 content of solid radwaste. The major source of production of C-14 in both PWR`s and BWRs was the activation of 0-17 in the water molecule and of N-14 dissolved in reactor coolant. The production of C-14 was estimated to range from 7 Ci/GW(e)-year to 11 Ci/GW(e)-year. The estimated range of the quantity of C-14 in LLWmore » was 1-2 Ci/ reactor-year which compares favorably with data obtained from shipping manifests. The environmental behavior of C-14 associated with low-level waste (LLW) disposal is greatly dependent upon its chemical speciation. This scoping study was performed to help identify the occurrence of inorganic and organic forms of C-14 in reactor coolant water and in primary coolant demineralization resins. These represent the major source for C-14 in LLW from nuclear power stations. Also, the behavior of inorganic and two of the organic forms of C-14 on soil uptake was determined by measuring distribution coefficients (Kd`s) on two soil types and a cement, using two different groundwater types. This study confirms that C-14 concentrations are significantly higher in the primary coolant from PWR stations compared to BWR stations. The C-14 followed trends of Co-60 generation during primary coolant demineralization at all but one of the stations examined. However, the C-14/Co-60 activity ratios measured by this study in resin samples through which samples of coolant were drawn were about 8 to 42 times higher than those reported for waste samples in the industry data base for PWR stations, and 15 to 730 times lower for the BWR stations.« less

Subsurface Hybrid Power Options for Oil & Gas Production at Deep Ocean Sites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, J C; Haut, R; Jahn, G

2010-02-19

An investment in deep-sea (deep-ocean) hybrid power systems may enable certain off-shore oil and gas exploration and production. Advanced deep-ocean drilling and production operations, locally powered, may provide commercial access to oil and gas reserves otherwise inaccessible. Further, subsea generation of electrical power has the potential of featuring a low carbon output resulting in improved environmental conditions. Such technology therefore, enhances the energy security of the United States in a green and environmentally friendly manner. The objective of this study is to evaluate alternatives and recommend equipment to develop into hybrid energy conversion and storage systems for deep ocean operations.more » Such power systems will be located on the ocean floor and will be used to power offshore oil and gas exploration and production operations. Such power systems will be located on the oceans floor, and will be used to supply oil and gas exploration activities, as well as drilling operations required to harvest petroleum reserves. The following conceptual hybrid systems have been identified as candidates for powering sub-surface oil and gas production operations: (1) PWR = Pressurized-Water Nuclear Reactor + Lead-Acid Battery; (2) FC1 = Line for Surface O{sub 2} + Well Head Gas + Reformer + PEMFC + Lead-Acid & Li-Ion Batteries; (3) FC2 = Stored O2 + Well Head Gas + Reformer + Fuel Cell + Lead-Acid & Li-Ion Batteries; (4) SV1 = Submersible Vehicle + Stored O{sub 2} + Fuel Cell + Lead-Acid & Li-Ion Batteries; (5) SV2 = Submersible Vehicle + Stored O{sub 2} + Engine or Turbine + Lead-Acid & Li-Ion Batteries; (6) SV3 = Submersible Vehicle + Charge at Docking Station + ZEBRA & Li-Ion Batteries; (7) PWR TEG = PWR + Thermoelectric Generator + Lead-Acid Battery; (8) WELL TEG = Thermoelectric Generator + Well Head Waste Heat + Lead-Acid Battery; (9) GRID = Ocean Floor Electrical Grid + Lead-Acid Battery; and (10) DOC = Deep Ocean Current + Lead-Acid Battery.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Downar, Thomas

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISONmore » / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.« less

User Data Package - Energy-Efficient Windows and Window Coverings for Naval Housing

DTIC Science & Technology

1990-07-01

1765 33 Savannah 1819 32 Tucson 1800 32 Winslow 4782 35 Idaho Yuma 974 33 Boise 5809 44 Lewiston 5542 46 Arkansas Pocatello 7033 43 Fort Smith 3292 35...Alexandria, VA DODDS Pac, FAC, Okinawa, Japan DOE Fed Energy Mgt Program, Wash, DC: INEL Tech Lib Reports Sta. Idaho Falls. ID; Knolls Atomic Pwr Lab

77 FR 9960 - Final Environmental Impact Statement for Extension of F-Line Streetcar Service to Fort Mason...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-21

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-1103-8840; 2051-P580-579] Final... AGENCY: National Park Service, Department of the Interior. ACTION: Notice of Availability of the Final... resources. Many also suggested various design ideas and other measures to help reduce these impacts. In...

78 FR 38359 - Approval of Record of Decision for Relocation of Cattle Point Road, San Juan Island National...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-26

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-12863; PPPWSAJHA0 PPMPSAS1Z.Y00000... Park, San Juan County, Washington AGENCY: National Park Service, Interior. ACTION: Notice of Record of....2), the Department of the Interior, National Park Service has prepared and approved a Record of...

77 FR 31655 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-29

... controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the... against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR [Pressurized... Institute] 97-06, Revision 3, ``Steam Generator Program Guidelines'' (Reference 1) and RG 1.121, ``Bases for...

Manufacturing Methods and Technology Measure for Fabrication of Silicon Transcalent Rectifier.

DTIC Science & Technology

1980-09-01

Prod Test/Eval’, z HA Kotler a Patent- Power & E 1 RM Roderick Env. Eng. & Test 1 JB Grosh Iron Mouptain - .l TUBE PARTS MFG. 5 RL SPALDING...AFAL/PODI ATTN: Working Group on Pwr. Devices (Mr. Philip Herron) 201 Varick Street Wright Patterson AFB, OH 45433 New York, NY 10014 Commander Mr

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

A study of life prediction differences for a nickel-base Alloy 690 using a threshold and a non-threshold model

NASA Astrophysics Data System (ADS)

Young, B. A.; Gao, Xiaosheng; Srivatsan, T. S.

2009-10-01

In this paper we compare and contrast the crack growth rate of a nickel-base superalloy (Alloy 690) in the Pressurized Water Reactor (PWR) environment. Over the last few years, a preponderance of test data has been gathered on both Alloy 690 thick plate and Alloy 690 tubing. The original model, essentially based on a small data set for thick plate, compensated for temperature, load ratio and stress-intensity range but did not compensate for the fatigue threshold of the material. As additional test data on both plate and tube product became available the model was gradually revised to account for threshold properties. Both the original and revised models generated acceptable results for data that were above 1 × 10 -11 m/s. However, the test data at the lower growth rates were over-predicted by the non-threshold model. Since the original model did not take the fatigue threshold into account, this model predicted no operating stress below which the material would effectively undergo fatigue crack growth. Because of an over-prediction of the growth rate below 1 × 10 -11 m/s, due to a combination of low stress, small crack size and long rise-time, the model in general leads to an under-prediction of the total available life of the components.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

CECE: Expanding the Envelope of Deep Throttling Technology in Liquid Oxygen/Liquid Hydrogen Rocket Engines for NASA Exploration Missions

NASA Technical Reports Server (NTRS)

Giuliano, Victor J.; Leonard, Timothy G.; Lyda, Randy T.; Kim, Tony S.

2010-01-01

As one of the first technology development programs awarded by NASA under the Vision for Space Exploration, the Pratt & Whitney Rocketdyne (PWR) Deep Throttling, Common Extensible Cryogenic Engine (CECE) program was selected by NASA in November 2004 to begin technology development and demonstration toward a deep throttling, cryogenic engine supporting ongoing trade studies for NASA s Lunar Lander descent stage. The CECE program leverages the maturity and previous investment of a flight-proven hydrogen/oxygen expander cycle engine, the PWR RL10, to develop and demonstrate an unprecedented combination of reliability, safety, durability, throttlability, and restart capabilities in high-energy, cryogenic, in-space propulsion. The testbed selected for the deep throttling demonstration phases of this program was a minimally modified RL10 engine, allowing for maximum current production engine commonality and extensibility with minimum program cost. Four series of demonstrator engine tests have been successfully completed between April 2006 and April 2010, accumulating 7,436 seconds of hot fire time over 47 separate tests. While the first two test series explored low power combustion (chug) and system instabilities, the third test series investigated and was ultimately successful in demonstrating several mitigating technologies for these instabilities and achieved a stable throttling ratio of 13:1. The fourth test series significantly expanded the engine s operability envelope by successfully demonstrating a closed-loop control system and extensive transient modeling to enable lower power engine starting, faster throttle ramp rates, and mission-specific ignition testing. The final hot fire test demonstrated a chug-free, minimum power level of 5.9%, corresponding to an overall 17.6:1 throttling ratio achieved. In total, these tests have provided an early technology demonstration of an enabling cryogenic propulsion concept with invaluable system-level technology data acquisition toward design and development risk mitigation for future lander descent main engines.

76 FR 23577 - Combined Notice of Filings #1

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-27

... Interconnection, L.L.C. submits tariff filing per 35.13(a)(2)(iii: Queue No. W3-124--Original Service Agreement No... Transmission Agreement with Auburndale Pwr Partners to be effective 5/1/2011. Filed Date: 04/21/2011. Accession... tariff filing per 35.13(a)(1): 04--21--11 Paris Rate Schedule 407 Settlement to be effective 3/31/2011...

Steam generators regulatory practices and issues in Spain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mendoza, C.; Castelao, C.; Ruiz-Colino, J.

1997-02-01

This paper presents the actual status of Spanish Steam Generator tubes, actions developed by PWR plant owners and submitted to CSN, and regulatory activities related to tube degradation mechanisms analysis; NDT tube inspection techniques; tube, tubesheet and TSPs integrity studies; tube plugging/repair criteria; preventive and corrective measures including whole SGs replacement; tube leak measurement methods and other operational aspects.

Organometallics in High Energy Chemistry.

DTIC Science & Technology

1983-10-31

Luines Physeical ftaenc Chemistry DepatneWu. SJI International. Menlo PWr *. CaiOwrnia M10 Rceived Nouvber 8. 1IM The otslytic formation of6nw carbon...support the idea that the metalloazocyclopropane intermediate is the reactive intermediate that leads to transalkylation. A discussion of the...exceptionally good correlation between the catalytic reactivity patterns of palladium black in its reactions with tertiary amines and those of homogeneous

Effect of surface state on the oxidation behavior of welded 308L in simulated nominal primary water of PWR

NASA Astrophysics Data System (ADS)

Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu; Han, En-Hou; Ke, Wei

2015-05-01

The oxidation behavior of 308L weld metal (WM) with different surface state in the simulated nominal primary water of pressurized water reactor (PWR) was studied by scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), X-ray diffraction (XRD) analyzer and X-ray photoelectron spectroscopy (XPS). After 480 h immersion, a duplex oxide film composed of a Fe-rich outer layer (Fe3O4, Fe2O3 and a small amount of NiFe2O4, Ni(OH)2, Cr(OH)3 and (Ni, Fe)Cr2O4) and a Cr-rich inner layer (FeCr2O4 and NiCr2O4) can be formed on the 308L WM samples with different surface state. The surface state has no influence on the phase composition of the oxide films but obviously affects the thickness of the oxide films and the morphology of the oxides (number & size). With increasing the density of dislocations and subgrain boundaries in the cold-worked superficial layer, the thickness of the oxide film, the number and size of the oxides decrease.

CHF considerations for highly moderated 100% MOX fuels PWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saphier, D.; Raymond, P.

1995-09-01

A feasibility study on using 100% MOX fuel in a PWR with increased moderating ratio, RMA, was initiated. In the proposed design all the parameters were chosen identical to the French 1450MW PWR, except the fuel pin diameter which was reduced to achieve higher moderating ratios, V{sub M}/V{sub F}, where V{sub M} and V{sub F} are the moderator and fuel volume respectively. Moderating ratios from 2 to 4 were considered. In the present study the thermal-hydraulic feasibility of using fuel assemblies with smaller diameter fuel pins was investigated. The major design constrain in this study was the critical heat fluxmore » (CHF). In order to maintain the fuel pin integrity under nominal operating and transient conditions, the minimum DNBR, (Departure from Nucleate Boiling Ratio given by CHF/q{close_quotes}{sub local}, where q{close_quotes}{sub local} is the local heat flux), has to be above a given value. The limitations of the existing CHF correlations for the present study are outlined. Two designs based on the conventional 17x17 fuel assembly and on the advanced 19x19 assembly meeting the MDNBR criteria and satisfying the control margin requirements, are proposed.« less

Development of new UV-I. I. Cerenkov Viewing Device

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki; Nemoto, Koshichi

1994-02-01

The Cerenkov glow images from boiling-water reactors (BWR) and pressurized-water reactors (PWR) irradiated fuel assemblies are generally used for inspections. However, sometimes it is difficult or impossible to identify the image by the conventional Cerenkov Viewing Device (CVD), because of the long cooling time and/or low burnup. Now a new UV-I.I. (Ultra-Violet light Image Intensifier) CVD has been developed, which can detect the very weak Cerenkov glow from spent fuel assemblies. As this new device uses the newly developed proximity focused type UV-I.I., Cerenkov photons are used efficiently, producing better quality Cerenkov glow images. Moreover, since the image is convertedmore » to a video signal, it is easy to improve the signal to noise ratio (S/N) by an image processor. The new CVD was tested at BWR and PWR power plants in Japan, with fuel burnups ranging from 6,200--33,000 MWD/MTU (megawatt days per metric ton of uranium) and cooling times ranging from 370 to 6,200 d. The tests showed that the new CVD is superior to the conventional STA/CRIEPI CVD, and could detect very feeble Cerenkov glow images using an image processor.« less

Crack opening area estimates in pressurized through-wall cracked elbows under bending

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franco, C.; Gilles, P.; Pignol, M.

1997-04-01

One of the most important aspects in the leak-before-break approach is the estimation of the crack opening area corresponding to potential through-wall cracks at critical locations during plant operation. In order to provide a reasonable lower bound to the leak area under such loading conditions, numerous experimental and numerical programs have been developed in USA, U.K. and FRG and widely discussed in literature. This paper aims to extend these investigations on a class of pipe elbows characteristic of PWR main coolant piping. The paper is divided in three main parts. First, a new simplified estimation scheme for leakage area ismore » described, based on the reference stress method. This approach mainly developed in U.K. and more recently in France provides a convenient way to account for the non-linear behavior of the material. Second, the method is carried out for circumferential through-wall cracks located in PWR elbows subjected to internal pressure. Finite element crack area results are presented and comparisons are made with our predictions. Finally, in the third part, the discussion is extended to elbows under combined pressure and in plane bending moment.« less

Joining dissimilar stainless steels for pressure vessel components

NASA Astrophysics Data System (ADS)

Sun, Zheng; Han, Huai-Yue

1994-03-01

A series of studies was carried out to examine the weldability and properties of dissimilar steel joints between martensitic and austenitic stainless steels - F6NM (OCr13Ni4Mo) and AISI 347, respectively. Such joints are important parts in, e.g. the primary circuit of a pressurized water reactor (PWR). This kind of joint requires both good mechanical properties, corrosion resistance and a stable magnetic permeability besides good weldability. The weldability tests included weld thermal simulation of the martensitic steel for investigating the influence of weld thermal cycles and post-weld heat treatment (PWHT) on the mechanical properties of the heat-affected zone (HAZ); implant testing for examining the tendency for cold cracking of martensitic steel; rigid restraint testing for determining hot crack susceptibility of the multi-pass dissimilar steel joints. The joints were subjected to various mechanical tests including a tensile test, bending test and impact test at various temperatures, as well as slow strain-rate test for examining the stress corrosion cracking tendency in the simulated environment of a primary circuit of a PWR. The results of various tests indicated that the quality of the tube/tube joints is satisfactory for meeting all the design requirements.

Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

NASA Astrophysics Data System (ADS)

Chen, Junjie; Xiao, Qian; Lu, Zhanpeng; Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan

2017-06-01

The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters.

On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

PubMed

Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

2016-08-01

For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

PRESSURIZED WATER REACTOR (PWR) PROJECT TECHNICAL PROGRESS REPORT FOR THE PERIOD DECEMBER 24, 1959 TO FEBRUARY 23, 1960

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

< 9 A < 2 6 < 7 4 8 9 6 2 6 equalizing vent valves on air locks 2, 4, and 5 was completed. An evaluation of the failed main coolant pump No. 1-80-F-737 was completed. The design for installing combination ball check and manual stop valves on the boiler water level sight glasses, to prevent the escape of steam should a defective sight glass develop, was completed. The main coolant pumps No. 80 and No. 79 were modified by increasing the radial clearance of the impeller wear ring and by removing the upper labyrinth ring. A designmore » for relocating the cooling water flow orifice 17-J4-17 was completed. Metallurgy: Preliminary data from the Bett 69-1 in-pile thermal conductivity capsules indicate that the thermal conductivity of as-sintered ZrO/sub 2/ 34 wt.% UO/sub 2/ appears to decrease from an initial value of about 1.6 Btu/hr-ft- deg F to about 0.7 Btu/hr-ft- deg F after 17 days irradiation in an estimated perturbed flux of 4 x 10/sup 13/. The thermal conductivities of UO/sub 2/ and BeO 51 wt.% UO/sub 2/ fuel remained unchanged during this time. Examination of the two failed X-3-1 fuel plates and the two failed CR-V-m fuel plates showed that a definite burnup limitation exists for bulk UO/sub 2/i of about 16 x 10/sup 20/ to 21.5 x 10/sup 20/ fissions/cc at which point the fuel increases in volume about 4- -5%. Irradiation of both fine and coarse dis-persions of 28 wt.% UO/sub 2/in BeO to exposures of about 11 x 10/sup 20/ fissions/cc shows this material has very poor dimensional stabllity and poor fission gas retention ability. The fine particles dispersion showed approximately 4.8 times the thickness increase as did the coarse particles. Interim examination of a bulk B/sub 4/ burnable poison plate irradiated in the HB-1 loop to about 60 at.% B/sup 10/ burnup showed a 17% increase in plate thickness. The technical feasibility of fabricating blanket receptacles with full length fuel channels and an integral cover plate by form rolling was established. Hack-pressure-bonding appears to be a suitable means of incorporating void volume in fuel compartments of oxide plates. High density (99% T.D.) and improved microstructure of B/sub 4/C-SiC burnable poisons are achieved when small (2 micron) B/sub 4/C particle size powder is used ia hot pressing compacts. Measurements of the self-diffusion coefficients of uranium in UO/sub 2/ by the method of surface activity decrease were completed. Experiments on the diffusion of Xe/sup 133/ in Core 2--type UO/sup 2/ fuel platelets were completed. Diffusion anaeals carried out at 1000 deg C on samples from the X-3-1 and the 14-28 irradiation tests show that the apparent diffusion coefficient for Kr/sup 85/ incresses considerably with burnup. An average activation energy for thoron emanation in UO/sub 2/ was estimated to be 44 kcal/mole. An initial experiment on the release of helium from slightly irradiated B/sub 4/C at 900 deg C resulted in a diffusion coefficient for helium of 3.5 x 10/sup -8/ Physics: Calculatad values for seed-blanket power sharing as a function of PWR-1 Seed 1 life were compared with measured data obtained from thermal instrumentation at Shippingport. Two-dimensional depletion studies in the PWR-2 "composite cell" geometry were completed for seed assembly configurations having different radial fuel zoning. An eighth core representation is being employed for a two- dimensional depletion calculation of PWR-2. An analysis of the effect on the axial power distribution of the nonuniform temperature distribution in an 8 ft PWR-2 core loaded with 295 kg of U/sup 235/ indicated that local variations in power density of as much as 15% may occur, relative to the distribution that would exist if the axial temperature distribution were uniform. A technique was developed which makes possible an approximately correct description of the neutron capture rate within small rectangular boron wafers in diffusion theory calculations. Seed peaking factors measured in a five-cluster slab of PWR-2 mock- up materials were measured and compared with calculated peaking factors obtained using the nuclear« less

Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, Randy K.; Odette, George R.; Almirall, Nathan

2017-05-01

The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughnessmore » loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.« less

Visualization of Flow in Pressurizer Spray Line Piping and Estimation of Thermal Stress Fluctuation Caused by Swaying of Water Surface

NASA Astrophysics Data System (ADS)

Oumaya, Toru; Nakamura, Akira; Onojima, Daisuke; Takenaka, Nobuyuki

The pressurizer spray line of PWR plants cools reactor coolant by injecting water into pressurizer. Since the continuous spray flow rate during commercial operation of the plant is considered insufficient to fill the pipe completely, there is a concern that a water surface exists in the pipe and may periodically sway. In order to identify the flow regimes in spray line piping and assess their impact on pipe structure, a flow visualization experiment was conducted. In the experiment, air was used substituted for steam to simulate the gas phase of the pressurizer, and the flow instability causing swaying without condensation was investigated. With a full-scale mock-up made of acrylic, flow under room temperature and atmospheric pressure conditions was visualized, and possible flow regimes were identified based on the results of the experiment. Three representative patterns of swaying of water surface were assumed, and the range of thermal stress fluctuation, when the surface swayed instantaneously, was calculated. With the three patterns of swaying assumed based on the visualization experiment, it was confirmed that the thermal stress amplitude would not exceed the fatigue endurance limit prescribed in the Japanese Design and Construction Code.

10 CFR 50.75 - Reporting and recordkeeping for decommissioning planning.

Code of Federal Regulations, 2010 CFR

2010-01-01

... investing or otherwise, that a prudent investor would use in the same circumstances. The term “prudent... than or equal to 3400 MWt $105 between 1200 MWt and 3400 MWt (For a PWR of less than 1200 MWt, use P... 3400 MWt (For a BWR of less than 1200 MWt, use P=1200 MWt) $(104+0.009P) (2) An adjustment factor at...

Experimental and Analytical Development of the Application of a Transit Laser Velocimeter

DTIC Science & Technology

1980-11-01

from 0.005" brass shim stock with carefully finished edges and chemically blackened surface and is slightly adjustable in position to compensate for...personnel by Mr. T. V. C i e l , ETF . Mr. V i r g i l Cline, PWr, and t h e f a c i l i t y technic ians . A 1" diameter underexpanded unheated

40 CFR 59.507 - What are the labeling requirements for aerosol coatings?

Code of Federal Regulations, 2010 CFR

2010-07-01

... in Table 1 of this subpart or a company-specific code, if that code is explained as required by § 59.511(a); (2) The applicable PWR limit for the product specified in Table 1 of this subpart; (3) The day... this subpart. (b) The label on the product must be displayed in such a manner that it is readily...

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vidal, Jean-Marc; Eschbach, Romain; Launay, Agnes

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of coolingmore » time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the CEA reference calculation codes for neutron physics with the JEFF-3.1.1 nuclear data set. (authors)« less

The benefits of a fast reactor closed fuel cycle in the UK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

Development of Hplc Techniques for the Analysis of Trace Metal Species in the Primary Coolant of a Pressurised Water Reactor.

NASA Astrophysics Data System (ADS)

Barron, Keiron Robert Philip

Available from UMI in association with The British Library. The need to monitor corrosion products in the primary circuit of a pressurised water reactor (PWR), at a concentration of 10pg ml^{-1} is discussed. A review of trace and ultra-trace metal analysis, relevant to the specific requirements imposed by primary coolant chemistry, indicated that high performance liquid chromatography (HPLC), coupled with preconcentration of sample was an ideal technique. A HPLC system was developed to determine trace metal species in simulated PWR primary coolant. In order to achieve the desired detection limit an on-line preconcentration system had to be developed. Separations were performed on Aminex A9 and Benson BC-X10 analytical columns. Detection was by post column reaction with Eriochrome Black T and Calmagite Linear calibrations of 2.5-100ng of cobalt (the main species of interest), were achieved using up to 200ml samples. The detection limit for a 200ml sample was 10pg ml^{-1}. In order to achieve the desired aim of on-line collection of species at 300^circ C, the use of inorganic ion-exchangers is essential. A novel application, utilising the attractive features of the inorganic ion-exchangers titanium dioxide, zirconium dioxide, zirconium arsenophosphate and pore controlled glass beads, was developed for the preconcentration of trace metal species at temperature and pressure. The performance of these exchangers, at ambient and 300^ circC was assessed by their inclusion in the developed analytical system and by the use of radioisotopes. The particular emphasis during the development has been upon accuracy, reproducibility of recovery, stability of reagents and system contamination, studied by the use of radioisotopes and response to post column reagents. This study in conjunction with work carried out at Winfrith, resulted in a monitoring system that could follow changes in coolant chemistry, on deposition and release of metal species in simulated PWR water loops. On -line detection of cobalt at 11pg ml^{ -1} was recorded, something which previously could not be performed by other techniques.

Fusion Helmet: Electronic Analysis

DTIC Science & Technology

2014-04-01

Table 1: LYR203-101B Board Feature P1 (SEC MODULE) DM648 GPIO PORn Video Ports (2) Bootmode SPI/UART I2C CLKIN MDIO DDR2 128MB/16bit SPI Flash 16...McASP EMAC-SGMII /2 MDIO I2C GPIO DDR2 128MB/16bit JTAG Memory CLKGEN I2C PGoodPGood PORn Pwr LED Power DSP SPI/UART DSP SPI/UARTSPI/UART Video Display

PWR PRELIMINARY DESIGN FOR PL-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Humphries, G. E.

1962-02-28

The pressurized water reactor preliminary design, the preferred design developed under Phase I of the PL-3 contract, is presented. Plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentatlon description, shielding description, auxiliary systems, power plant equipment, waste dispusal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information, and a training program outline are given. (auth)

Solar Photovoltaic and Liquid Natural Gas Opportunities for Command Naval Region Hawaii

DTIC Science & Technology

2014-12-01

Utilities Commission xii PV Photovoltaic Pwr Power RE Renewable Energy Re-gas Regasification RFP Request For Proposal RMI Rocky... forecasted LS diesel price and the forecasted LNG delivered-to-the- power -plant cost. The forecast for LS diesel by FGE from year 2020–2030 is seen...annual/html/epa_08_01.html Electric Power Research Institute. (July, 2010). Addressing solar photovoltaic operations and maintenance challenges: A

Two-phase Fluid Selection for High-temperature Automotive Platforms

DTIC Science & Technology

2012-09-01

cases, extra work by researchers can be done to uprate the device, either by parameter conformance, parameter re-characterization, or stress balancing...chemical reactivity , noble metal oxidization, intermetallic growth, CTE mismatch and other failure mechanisms are an issue with wide bandgap...21005-1852 1 US ARMY TARDEC ATTN AMSRD TAR E/ PWR C SPANGLER 6501 E 11 MILE RD, BLDG 212 WARREN MI 48397-5000 1 UNIV OF MARYLAND

CH-47C Vulnerability Reduction Modification Program - Fly-by-Wire Backup Demonstration

DTIC Science & Technology

1976-08-01

Actuator Position for Combined Axis Input ............................. 91 4 Systems Assessment Summary................... 95 C-1 Instrumentation Parameters ...SERVO CARD jEETO FROM MIXERS SUfEV __________ HYLIC AMPL AMPLVLE SHUT-O- DOWN DC PWR LOGIC REA MIONITOR SUMMER *O:EO SWITCH- BUFFER OVER 1 NETWORK...and ranels (Figures 12 and 13). The existing DELS preflight test set, which provides access to the system parameters , was installed along with the

Fretting wear behaviors of a dual-cooled nuclear fuel rod under a simulated rod vibration

NASA Astrophysics Data System (ADS)

Lee, Young-Ho; Kim, Hyung-Kyu; Kang, Heung-Seok; Yoon, Kyung-Ho; Kim, Jae-Yong; Lee, Kang-Hee

2012-06-01

Recently, a dual-cooled fuel (i.e., annular fuel) that is compatible with current operating PWR plants has been proposed in order to realize both a considerable amount of power uprating and an increase of safety margins. As the design concept should be compatible with current operating PWR plants, however, it shows a narrow gap between the fuel rods when compared with current solid nuclear fuel arrays and needs to modify the spacer grid shapes and their positions. In this study, fretting wear tests have been performed to evaluate the wear resistance of a dual-cooled fuel by using a proposed spring and dimple of spacer grids that have a cantilever type and hemispherical shape, respectively. As a result, the wear volume of the spring specimen gradually increases as the contact condition is changed from a certain gap, just contact to positive force. However, in the dimple specimen, just contact condition shows a large wear volume. In addition, a circular rod motion at upper region of contact surface is gradually increased and its diametric size depends on the wear depth increase. Based on the test results, the fretting wear resistance of the proposed spring and dimple is analyzed by comparing the wear measurement results and rod motion in detail.

Current training initiatives at Nuclear Electric plc

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fowler, C.D.

1993-01-01

Nuclear Electric, one of the three generating companies to emerge from the demise of the U.K.'s Central Electricity Generating Board (CEGB), owns and operates the commercial nuclear power stations in England and Wales. The U.K. government proscribed further construction beyond Sizewell B, the United Kingdom's first pressurized water reactor (PWR) station, pending the outcome of a review of the future of nuclear power to be held in 1994. The major challenges facing Nuclear Electric at its formation in 1990 were therefore to demonstrate that nuclear power is safe, economical, and environmentally acceptable and to complete the PWR station under constructionmore » on time and within budget. A significant number of activities were started that were designed to increase output, reduce costs, and ensure that the previous excellent safety standards were maintained. A major activity was to reduce the numbers of staff employed, with a recognition from the outset that this reduction could only be achieved with a significant human resource development program. Future company staff would have to be competent in more areas and more productive. This paper summarizes some of the initiatives currently being pursued throughout the company and the progress toward ensuring that staff with the required competences are available to commission and operate the Sizewell B program in 1994.« less

System Engineering for J-2X Development: The Simpler, the Better

NASA Technical Reports Server (NTRS)

Kelly, William M.; Greasley, Paul; Greene, William D.; Ackerman, Peter

2008-01-01

The Ares I and Ares V Vehicles will utilize the J-2X rocket engine developed for NASA by the Pratt and Whitney Rocketdyne Company (PWR) as the upper stage engine (USE). The J-2X is an improved higher power version of the original J-2 engine used for Apollo. System Engineering (SE) facilitates direct and open discussions of issues and problems. This simple idea is often overlooked in large, complex engineering development programs. Definition and distribution of requirements from the engine level to the component level is controlled by Allocation Reports which breaks down numerical design objectives (weight, reliability, etc.) into quanta goals for each component area. Linked databases of design and verification requirements help eliminate redundancy and potential mistakes inherent in separated systems. Another tool, the Architecture Design Description (ADD), is used to control J-2X system architecture and effectively communicate configuration changes to those involved in the design process. But the proof of an effective process is in successful program accomplishment. SE is the methodology being used to meet the challenge of completing J-2X engine certification 2 years ahead of any engine program ever developed at PWR. This paper describes the simple, better SE tools and techniques used to achieve this success.

Investigation into the effect of water chemistry on corrosion product formation in areas of accelerated flow

NASA Astrophysics Data System (ADS)

McGrady, John; Scenini, Fabio; Duff, Jonathan; Stevens, Nicholas; Cassineri, Stefano; Curioni, Michele; Banks, Andrew

2017-09-01

The deposition of CRUD (Chalk River Unidentified Deposit) in the primary circuit of a Pressurised Water Reactor (PWR) is known to preferentially occur in regions of the circuit where flow acceleration of coolant occurs. A micro-fluidic flow cell was used to recreate accelerated flow under simulated PWR conditions, by flowing water through a disc with a central micro-orifice. CRUD deposition was reproduced on the disc, and CRUD Build-Up Rates (BUR) in various regions of the disc were analysed. The effect of the local environment on BUR was investigated. In particular, the effect of flow velocity, specimen material and Fe concentration were considered. The morphology and composition of the deposits were analysed with respect to experimental conditions. The BUR of CRUD was found to be sensitive to flow velocity and Fe concentration, suggesting that mass transfer is an important factor. The morphology of the deposit was affected by the specimen material indicating a dependence on surface/particle electrostatics meaning surface chemistry plays an important role in deposition. The preferential deposition of CRUD in accelerated flow regions due to electrokinetic effects was observed and it was shown that higher Fe concentrations in solution increased BURs within the orifice whereas increased flow velocity reduced BURs.

Penetrative Internal Oxidation from Alloy 690 Surfaces and Stress Corrosion Crack Walls during Exposure to PWR Primary Water

NASA Astrophysics Data System (ADS)

Olszta, Matthew J.; Schreiber, Daniel K.; Thomas, Larry E.; Bruemmer, Stephen M.

Analytical electron microscopy and three-dimensional atom probe tomography (ATP) examinations of surface and near-surface oxidation have been performed on Ni-30%Cr alloy 690 materials after exposure to high-temperature, simulated PWR primary water. The oxidation nanostructures have been characterized at crack walls after stress-corrosion crack growth tests and at polished surfaces of unstressed specimens for the same alloys. Localized oxidation was discovered for both crack walls and surfaces as continuous filaments (typically <10 nm in diameter) extending from the water interface into the alloy 690 matrix reaching depths of 500 nm. These filaments consisted of discrete, plate-shaped Cr2O3 particles surrounded by a distribution of nanocrystalline, rock-salt (Ni-Cr-Fe) oxide. The oxide-containing filament depth was found to increase with exposure time and, at longer times, the filaments became very dense at the surface leaving only isolated islands of metal. Individual dislocations were oxidized in non-deformed materials, while the oxidation path appeared to be along more complex dislocation substructures in heavily deformed materials. This paper will highlight the use of high resolution scanning and transmission electron microscopy in combination with APT to better elucidate the microstructure and microchemistry of the filamentary oxidation.

TRIGA MARK-II source term

DOE Office of Scientific and Technical Information (OSTI.GOV)

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

Development Status of the CECE Cryogenic Deep Throttling Demonstrator Engine

NASA Technical Reports Server (NTRS)

2008-01-01

As one of the first technology development programs awarded by NASA under the U.S. Space Exploration Policy (USSEP), the Pratt & Whitney Rocketdyne (PWR) Deep Throttling, Common Extensible Cryogenic Engine (CECE) program was selected by NASA in November 2004 to begin technology development and demonstration toward a deep throttling, cryogenic engine supporting ongoing trade studies for NASA's Lunar Lander descent stage. The CECE program leverages the maturity and previous investment of a flight-proven hydrogen/oxygen expander cycle engine, the PWR RLI0, to develop and demonstrate an unprecedented combination of reliability, safety, durability, throttlability, and restart capabilities in a high-energy, cryogenic engine. The testbed selected for the deep throttling demonstration phases of this program was a minimally modified RL10 engine, allowing for maximum current production engine commonality and extensibility with minimum program cost. Two series of demonstrator engine tests, the first in April-May 2006 and the second in March-April 2007, have demonstrated in excess of 10:1 throttling of the hydrogen/oxygen expander cycle engine. Both test series have explored a combustion instability ("chug") environment at low throttled power levels. These tests have provided an early demonstration of an enabling cryogenic propulsion concept with invaluable system-level technology data acquisition toward design and development risk mitigation for future CECE Demonstrator engine tests.

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

Fatigue limit and Hysteresis Behavior of Type 304L Stainless Steel in Air and PWR Water, at 150°C and 300°C

NASA Astrophysics Data System (ADS)

Solomon, H. D.; Amzallag, C.; Vallee, A. J.; DeLair, R. E.

This is a study of the 107 cycle fatigue limit of Type 304L Stainless Steel, as measured in fully reversed (R=-1) load-controlled tests, at 150°C and 300°C, in air and PWR water. The staircase method was used to determine the fatigue limit. The tests run here utilized a cycle frequency of 1.818Hz and are compared to other tests from the literature that were run at 30Hz. The fatigue limit measured in the tests run at the high frequency was higher than that measured here. This is explained by measurements of the strain developed during cycling, using the different cycle frequencies. The tests run at the higher frequencies yielded lower strains for a given stress and, as expected, this resulted in higher fatigue limits. Using 107 cycles to define a run-out also led to a lower fatigue limit. These results are important as most previous fatigue limit measurements utilized 106 cycles or less to define a run-out, and when lives as long as 107 cycles are used the tests are generally run at high cycle frequencies, thus leading to higher fatigue limits than those measured here.

Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water

NASA Astrophysics Data System (ADS)

Cissé, Sarata; Tanguy, Benoit; Laffont, Lydia; Lafont, Marie-Christine; Guerre, Catherine; Andrieu, Eric

The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ɛp/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans <111>, that swept defects or γ' precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water

NASA Astrophysics Data System (ADS)

Fournier, L.; Savoie, M.; Delafosse, D.

2007-06-01

The low cycle fatigue (LCF) behaviour of precipitation-strengthened A-286 austenitic stainless steel was first investigated at room temperature under 0.2% plastic strain control. LCF led to hardening for the first 20 cycles and then to significant softening. LCF-induced dislocation microstructure was characterized using both bright and dark-field imaging techniques in transmission electron microscopy. Cycling softening was correlated with the formation of precipitate-free localized deformation bands. The effect of these precipitate-free localized deformation bands on A-286 stress corrosion cracking (SCC) behaviour in PWR primary water was then examined by means of constant extension rate tensile (CERT) tests at 320 °C and 360 °C. Comparative CERT tests were performed on companion specimens with similar yield stress but pre-fatigued to a few cycles (4-8) or between 125 and 200 cycles. Specimens pre-fatigued to a few cycles with no precipitate-free localized deformation bands exhibited little susceptibility to intergranular SCC (IGSCC). In contrast, the presence of precipitate-free localized deformation bands formed by pre-fatigue to between 125 and 200 cycles strongly promoted IGSCC. The interest of the approach used in this study is to provide insight into the role of localized deformation in irradiation assisted stress corrosion cracking.

Characterization of ion irradiation effects on the microstructure, hardness, deformation and crack initiation behavior of austenitic stainless steel:Heavy ions vs protons

NASA Astrophysics Data System (ADS)

Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

2018-04-01

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a complex phenomenon of degradation which can have a significant influence on maintenance time and cost of core internals of a Pressurized Water Reactor (PWR). Hence, it is an issue of concern, especially in the context of lifetime extension of PWRs. Proton irradiation is generally used as a representative alternative of neutron irradiation to improve the current understanding of the mechanisms involved in IASCC. This study assesses the possibility of using heavy ions irradiation to evaluate IASCC mechanisms by comparing the irradiation induced modifications (in microstructure and mechanical properties) and cracking susceptibility of SA 304 L after both type of irradiations: Fe irradiation at 450 °C and proton irradiation at 350 °C. Irradiation-induced defects are characterized and quantified along with nano-hardness measurements, showing a correlation between irradiation hardening and density of Frank loops that is well captured by Orowan's formula. Both irradiations (iron and proton) increase the susceptibility of SA 304 L to intergranular cracking on subjection to Constant Extension Rate Tensile tests (CERT) in simulated nominal PWR primary water environment at 340 °C. For these conditions, cracking susceptibility is found to be quantitatively similar for both irradiations, despite significant differences in hardening and degree of localization.

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

Plasmon Spectroscopy Applied to Biomolecular Interactions in Membranes

NASA Astrophysics Data System (ADS)

Tollin, Gordon

2010-03-01

Plasmon-waveguide resonance (PWR) is an optical spectroscopy method that can provide information about materials immobilized on the surface of a plasmon resonator consisting of a right angle prism coated with thin layers of a metal (approx. 50 nm; usually silver) and a dielectric (approx. 500 nm; usually silica). The technique has been developed in our laboratory and is an extension of the more commonly used surface plasmon resonance (SPR) method, having higher sensitivity (20-50 fold) and resolution (10-20 fold). The dielectric layer allows plasmon excitation by light whose electric vector is polarized both perpendicular and parallel to the sensor surface, in contrast to SPR that can only utilize perpendicular polarized excitation. This allows both mass density and mass distribution to be characterized in uniaxially oriented deposited materials, such as biomembranes. We have utilized this technique to investigate binding interactions between membrane-incorporated protein receptors and their ligands (both proteins and small molecules), using both purified receptors inserted into lipid bilayers and membranes derived from cells expressing these receptors. Such studies have provided many new insights into biological signaling events. Inasmuch as many of these receptors are targets for approximately 50 percent of ethical drugs, PWR can be a useful methodology for drug discovery in the pharmaceutical industry. Examples of these experiments will be presented.

Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

NASA Astrophysics Data System (ADS)

Rouf; Su'ud, Zaki

2016-08-01

Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

Modeling of local steam condensation on walls in presence of non-condensable gases. Application to a loca calculation in reactor containment using the multidimensional geyser/tonus code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benet, L.V.; Caroli, C.; Cornet, P.

1995-09-01

This paper reports part of a study of possible severe pressurized water reactor (PWR) accidents. The need for containment modeling, and in particular for a hydrogen risk study, was reinforced in France after 1990, with the requirement that severe accidents must be taken into account in the design of future plants. This new need of assessing the transient local hydrogen concentration led to the development, in the Mechanical Engineering and Technology Department of the French Atomic Energy Commission (CEA/DMT), of the multidimensional code GEYSER/TONUS for containment analysis. A detailed example of the use of this code is presented. The mixturemore » consisted of noncondensable gases (air or air plus hydrogen) and water vapor and liquid water. This is described by a compressible homogeneous two-phase flow model and wall condensation is based on the Chilton-Colburn formula and the analogy between heat and mass transfer. Results are given for a transient two-dimensional axially-symmetric computation for the first hour of a simplified accident sequence. In this there was an initial injection of a large amount of water vapor followed by a smaller amount and by hydrogen injection.« less

Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Barua, Bipul; Soppet, William K.

This report provides an update of an earlier assessment of environmentally assisted fatigue for components in light water reactors. This report is a deliverable in September 2016 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2016 report, we presented a detailed thermal-mechanical stress analysis model for simulating the stress-strain state of a reactor pressure vessel and its nozzles under grid-load-following conditions. In this report, we provide stress-controlled fatigue test data for 508 LAS base metal alloy under different loading amplitudes (constant, variable, and random grid-load-following) and environmental conditions (in airmore » or pressurized water reactor coolant water at 300°C). Also presented is a cyclic plasticity-based analytical model that can simultaneously capture the amplitude and time dependency of the component behavior under fatigue loading. Results related to both amplitude-dependent and amplitude-independent parameters are presented. The validation results for the analytical/mechanistic model are discussed. This report provides guidance for estimating time-dependent, amplitude-independent parameters related to material behavior under different service conditions. The developed mechanistic models and the reported material parameters can be used to conduct more accurate fatigue and ratcheting evaluation of reactor components.« less

Conclusions and Recommendations Regarding the Deep Sea Hybrid Power Systems Initial Study

DTIC Science & Technology

2010-06-01

proton-exchange membrane fuel cells ( PEMFC ) powered with hydrogen and oxygen, similar to that used on proven subsurface vessels; (2) fuel-cells...AND STORAGE OPTIONS CONSIDERED FOR INITIAL STUDY NO. NOMENCLATURE DESCRIPTION 1 PWR Nuclear Reactor + Battery 2 FC1 PEMFC + Line for surface O2...Wellhead Gas + Reformer + Battery 3 FC2 PEMFC + Stored O2 + Wellhead Gas + Reformer + Battery 4 SV1 PEMFC + Submersible Vehicle for O2 Transport

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

Heat transfer during condensation of steam from steam-gas mixtures in the passive safety systems of nuclear power plants

NASA Astrophysics Data System (ADS)

Portnova, N. M.; Smirnov, Yu B.

2017-11-01

A theoretical model for calculation of heat transfer during condensation of multicomponent vapor-gas mixtures on vertical surfaces, based on film theory and heat and mass transfer analogy is proposed. Calculations were performed for the conditions implemented in experimental studies of heat transfer during condensation of steam-gas mixtures in the passive safety systems of PWR-type reactors of different designs. Calculated values of heat transfer coefficients for condensation of steam-air, steam-air-helium and steam-air-hydrogen mixtures at pressures of 0.2 to 0.6 MPa and of steam-nitrogen mixture at the pressures of 0.4 to 2.6 MPa were obtained. The composition of mixtures and vapor-to-surface temperature difference were varied within wide limits. Tube length ranged from 0.65 to 9.79m. The condensation of all steam-gas mixtures took place in a laminar-wave flow mode of condensate film and turbulent free convection in the diffusion boundary layer. The heat transfer coefficients obtained by calculation using the proposed model are in good agreement with the considered experimental data for both the binary and ternary mixtures.

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Consistent criticality and radiation studies of Swiss spent nuclear fuel: The CS2M approach.

PubMed

Rochman, D; Vasiliev, A; Ferroukhi, H; Pecchia, M

2018-06-15

In this paper, a new method is proposed to systematically calculate at the same time canister loading curves and radiation sources, based on the inventory information from an in-core fuel management system. As a demonstration, the isotopic contents of the assemblies come from a Swiss PWR, considering more than 6000 cases from 34 reactor cycles. The CS 2 M approach consists in combining four codes: CASMO and SIMULATE to extract the assembly characteristics (based on validated models), the SNF code for source emission and MCNP for criticality calculations for specific canister loadings. The considered cases cover enrichments from 1.9 to 5.0% for the UO 2 assemblies and 4.8% for the MOX, with assembly burnup values from 7 to 74 MWd/kgU. Because such a study is based on the individual fuel assembly history, it opens the possibility to optimize canister loadings from the point-of-view of criticality, decay heat and emission sources. Copyright © 2018 Elsevier B.V. All rights reserved.

Preliminary Thermal Modeling of HI-STORM 100 Storage Modules at Diablo Canyon Power Plant ISFSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuta, Judith M.; Adkins, Harold E.

Thermal analysis is being undertaken at Pacific Northwest National Laboratory (PNNL) in support of inspections of selected storage modules at various locations around the United States, as part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development. This report documents pre-inspection predictions of temperatures for two modules at the Diablo Canyon Power Plant ISFSI identified as candidates for inspection. These are HI-STORM 100 modules of a site-specific design for storing PWR 17x17 fuel in MPC-32 canisters. The temperature predictions reported in this document were obtained with detailedmore » COBRA-SFS models of these storage systems, with the following boundary conditions and assumptions. • storage module overpack configuration based on FSAR documentation of HI-STORM100S-218, Version B; due to unavailability of site-specific design data for Diablo Canyon ISFSI modules • Individual assembly and total decay heat loadings for each canister, based on at-loading values provided by PG&E, “aged” to time of inspection using ORIGEN modeling o Special Note: there is an inherent conservatism of unquantified magnitude – informally estimated as up to approximately 20% -- in the utility-supplied values for at-loading assembly decay heat values • Axial decay heat distributions based on a bounding generic profile for PWR fuel. • Axial location of beginning of fuel assumed same as WE 17x17 OFA fuel, due to unavailability of specific data for WE17x17 STD and WE 17x17 Vantage 5 fuel designs • Ambient conditions of still air at 50°F (10°C) assumed for base-case evaluations o Wind conditions at the Diablo Canyon site are unquantified, due to unavailability of site meteorological data o additional still-air evaluations performed at 70°F (21°C), 60°F (16°C), and 40°F (4°C), to cover a range of possible conditions at the time of the inspection. (Calculations were also performed at 80°F (27°C), for comparison with design basis assumptions.) All calculations are for steady-state conditions, on the assumption that the surfaces of the module that are accessible for temperature measurements during the inspection will tend to follow ambient temperature changes relatively closely. Comparisons to the results of the inspections, and post-inspection evaluations of temperature measurements obtained in the specific modules, will be documented in a separate follow-on report, to be issued in a timely manner after the inspection has been performed.« less

Effect of lifestyle interventions of pregnant women on their dietary habits, lifestyle behaviors, and weight gain: a randomized controlled trial.

PubMed

Aşcı, Özlem; Rathfisch, Gülay

2016-02-24

Although it is known that lifestyle behaviors of pregnant women are closely related to maternal and fetal health, number of data concerning efficacy of intervention on lifestyle during pregnancy is limited. The purpose of this study is to determine the effect of lifestyle interventions on improving dietary habits and lifestyle behaviors, ensuring gestational weight gain (GWG) within recommended levels and limiting postpartum weight retention (PWR). The study was conducted as a randomized controlled trial in a family health center located in Istanbul, Turkey, between June 2011 and July 2012. The primary outcomes were GWG, and the proportion of pregnant women whose GWG was within the Institute of Medicine (IOM) guidelines. One hundred two pregnant women with gestation ≤12 weeks, age ≥18 years, gravidity ≤2, and who did not intend to lose weight in prepregnancy period were randomly included in this study as intervention (n = 51) and control (n = 51) groups. The study was completed with 45 women for each group. The control group received routine antenatal care. The intervention group was received an individualized lifestyle intervention focusing on healthy lifestyle, diet, exercise, and weight monitoring as four sessions at 12-15, 16-18, 20-24, and 37 weeks gestation. Lifestyle behaviors were evaluated with Health-Promoting Lifestyle Profile-II. Dietary habits were assessed by 3-day dietary recalls, and weight was followed from pregnancy until 6 weeks postpartum. The lifestyle interventions had a significant effect on improving lifestyle behaviors, protein intake, percentage of energy from protein, calcium, magnesium, iron, zinc, and vegetable intakes when adjusted for confounders (p < 0.05). The proportion of women who were within the IOM recommendations was higher in the intervention group (51.1 %) than in the control group (28.9 %) The odds ratio for GWG within IOM was statistically significant between the groups (OR = 0.59, 95 % CI, 0.45-0.72). There were no difference between groups in terms of the other dietary intakes, total GWG, and PWR (p > 0.05). Lifestyle intervention improves the lifestyle behaviors during pregnancy and increases the appropriate GWG for prepregnancy body mass index (BMI), but it has a limited effect in terms of improving dietary habits and has no effect on PWR.

Coupled calculation of the radiological release and the thermal-hydraulic behavior of a 3-loop PWR after a SGTR by means of the code RELAP5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Hove, W.; Van Laeken, K.; Bartsoen, L.

1995-09-01

To enable a more realistic and accurate calculation of the radiological consequences of a SGTR, a fission product transport model was developed. As the radiological releases strongly depend on the thermal-hydraulic transient, the model was included in the RELAP5 input decks of the Belgian NPPs. This enables the coupled calculation of the thermal-hydraulic transient and the radiological release. The fission product transport model tracks the concentration of the fission products in the primary circuit, in each of the SGs as well as in the condenser. This leads to a system of 6 coupled, first order ordinary differential equations with timemore » dependent coefficients. Flashing, scrubbing, atomisation and dry out of the break flow are accounted for. Coupling with the thermal-hydraulic calculation and correct modelling of the break position enables an accurate calculation of the mixture level above the break. Pre- and post-accident spiking in the primary circuit are introduced. The transport times in the FW-system and the SG blowdown system are also taken into account, as is the decontaminating effect of the primary make-up system and of the SG blowdown system. Physical input parameters such as the partition coefficients, half life times and spiking coefficients are explicitly introduced so that the same model can be used for iodine, caesium and noble gases.« less

A Comparative Study of Various Electric Propulsion Systems and their Impact on a Nominal Ship Design

DTIC Science & Technology

1987-06-01

Permanent Magnet Motors ," Advanced Mechanical Technology, Inc., 1983. 20. Marshall, McMurray, Richter, Webster, and...December 1977. 21. Triezenbarg, Greene, Hannan, and Dvorsky, "Study of Permanent Magnet Motors for Naval Propulsion," Westinghcuse Research Report 80=9B2...34 Paper 71 CP 155-PWR, IEEE Winter Power Meeting, New York, February 1971. 34. Ireland, James R., Ceramic Permanent - Magnet Motors , McGraw-Hill, New York, 1968. 206

Automotive Test Rig Final Design Report. Volume 2. Control System.

DTIC Science & Technology

1986-01-01

Pressure Switch Status P27 Low Brake Release Pressure Switch Status P26 Low Brake...Supply Pressure Switch Status P25 Low Port Charge Pump Pressure Switch Status P24 Low Starboard Charge Pump Pressure Switch Status P23 Hydraulic Filter By...Sensed Switch Status P31 Low Scavenge Pump Pressure Switch Status P30 P37 Signal Return for Computer J21 Not Used J22 P A +24 B Pwr Rtn C Ground C

Nuclear Fuel Depletion Analysis Using Matlab Software

NASA Astrophysics Data System (ADS)

Faghihi, F.; Nematollahi, M. R.

Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.

Limited Artificial and Natural Icing Tests Production UH-60A Helicopter (Re-Evaluation).

DTIC Science & Technology

1981-08-01

parameters , and definitions of icing types and severities are presented in appendix D. 2 RESULTS AND DISCUSSION GENERAL 9. Artificial and natural icing flight...anti-ice off, the system may be reactivated by cycling the appropriate windshield anti-ice switch. The windshield anti-ice system is fully operational...is off, then the fault monitor illuminates the respective PWR light on its front panel. The light informs the crew that further action is requied to

A NIST Kinetic Data Base for PAH Reaction and Soot Particle Inception During Combusion

DTIC Science & Technology

2007-12-01

in Computational Fluid Dynamics (CFD) codes hat have lead to the capability of describing complex reactive flow problems and thus simulating... parameters . However in the absence of data estimates must be made. Since the chemistry of combustion is extremely complex and for proper description...118:381-389 9. Babushok, V. and Tsang, W., J. Prop. and Pwr . 20 (2004) 403-414. 10. . Fournet, R., Warth, V., Glaude, P.A., Battin-Leclerc, F

Costs, equity, efficiency and feasibility of identifying the poor in Ghana's National Health Insurance Scheme: empirical analysis of various strategies.

PubMed

Aryeetey, Genevieve Cecilia; Jehu-Appiah, Caroline; Spaan, Ernst; Agyepong, Irene; Baltussen, Rob

2012-01-01

To analyse the costs and evaluate the equity, efficiency and feasibility of four strategies to identify poor households for premium exemptions in Ghana's National Health Insurance Scheme (NHIS): means testing (MT), proxy means testing (PMT), participatory wealth ranking (PWR) and geographic targeting (GT) in urban, rural and semi-urban settings in Ghana. We conducted the study in 145-147 households per setting with MT as our gold standard strategy. We estimated total costs that included costs of household surveys and cost of premiums paid to the poor, efficiency (cost per poor person identified), equity (number of true poor excluded) and the administrative feasibility of implementation. The cost of exempting one poor individual ranged from US$15.87 to US$95.44; exclusion of the poor ranged between 0% and 73%. MT was most efficient and equitable in rural and urban settings with low-poverty incidence; GT was efficient and equitable in the semi-urban setting with high-poverty incidence. PMT and PWR were less equitable and inefficient although feasible in some settings. We recommend MT as optimal strategy in low-poverty urban and rural settings and GT as optimal strategy in high-poverty semi-urban setting. The study is relevant to other social and developmental programmes that require identification and exemptions of the poor in low-income countries. © 2011 Blackwell Publishing Ltd.

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

Korean standard nuclear plant ex-vessel neutron dosimetry program Ulchin 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duo, J.I.; Chen, J.; Kulesza, J.A.

2011-07-01

A comprehensive ex-vessel neutron dosimetry (EVND) surveillance program has been deployed in 16 pressurized water reactors (PWR) in South Korea and EVND dosimetry sets have already been installed and analyzed in Westinghouse reactor designs. In this paper, the unique features of the design, training, and installation in the Korean standard nuclear plant (KSNP) Ulchin Unit 4 are presented. Ulchin Unit 4 Cycle 9 represents the first dosimetry analyzed from the EVND design deployed in KSNP plants: Yonggwang Units 3 through 6 and Ulchin Units 3 through 6. KSNP's cavity configuration precludes a conventional installation from the cavity floor. The solution,more » requiring the installation crew to access the cavity at an elevation of the active core, places a premium on rapid installation due to high area dose rates. Numerous geometrical features warranted the use of a detailed design in true 3D mechanical design software to control interferences. A full-size training mockup maximized the crew ability to correctly install the instrument in minimum time. The analysis of the first dosimetry set shows good agreements between measurement and calculation within the associated uncertainties. A complete EVND system has been successfully designed, installed, and analyzed for a KNSP plant. Current and future EVND analyses will continue supporting the successful operation of PWR units in South Korea. (authors)« less

The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebak, R B; Hua, F H

2004-07-12

Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attemptmore » is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.« less

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Task related doses in Spanish pressurized water reactors over the period 1988-1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Donnell, P.; Labarta, T.; Amor, I.

1995-03-01

In order to evaluate in depth the collective dose trend and its correlation with the effectiveness of the practical application of the ALARA principle in Spanish nuclear facilities, and base the different policy lines to promote this criteria, the CSN has fullfilled an analysis of the task related doses data over the period 1988-1992. Previously, the CSN had required to the utilities the compilation of their refuelling outage collective dose from 1988 according with a predeterminate number of tasks, in order to have available a representative and retrospective set of data in an homogeneous way and coherent with the internationalmore » data banks on occupational exposure in NPP, as the CEC and the NEA ones. The scope of this analysis was the following: first, the collective dose summaries for outage tasks and departments for PWR and for BWR, including the minimum, maximum and average dose (and statistics data) for 18 different refuelling outage tasks and 12 personal departments for each generation of each type of rector, the task and department related collective dose trends in each plant and in each generation, and second, the dose reduction techniques having been used during that period in each plant and the relative level of adoption. In this presentation the main results and conclusions of the first part of the study are reviewed for PWR.« less

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles J.

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE PAGES

Youinou, Gilles J.

2017-05-04

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Constraints on silicates formation in the Si-Al-Fe system: Application to hard deposits in steam generators of PWR nuclear reactors

NASA Astrophysics Data System (ADS)

Berger, Gilles; Million-Picallion, Lisa; Lefevre, Grégory; Delaunay, Sophie

2015-04-01

Introduction: The hydrothermal crystallization of silicates phases in the Si-Al-Fe system may lead to industrial constraints that can be encountered in the nuclear industry in at least two contexts: the geological repository for nuclear wastes and the formation of hard sludges in the steam generator of the PWR nuclear plants. In the first situation, the chemical reactions between the Fe-canister and the surrounding clays have been extensively studied in laboratory [1-7] and pilot experiments [8]. These studies demonstrated that the high reactivity of metallic iron leads to the formation of Fe-silicates, berthierine like, in a wide range of temperature. By contrast, the formation of deposits in the steam generators of PWR plants, called hard sludges, is a newer and less studied issue which can affect the reactor performance. Experiments: We present here a preliminary set of experiments reproducing the formation of hard sludges under conditions representative of the steam generator of PWR power plant: 275°C, diluted solutions maintained at low potential by hydrazine addition and at alkaline pH by low concentrations of amines and ammoniac. Magnetite, a corrosion by-product of the secondary circuit, is the source of iron while aqueous Si and Al, the major impurities in this system, are supplied either as trace elements in the circulating solution or by addition of amorphous silica and alumina when considering confined zones. The fluid chemistry is monitored by sampling aliquots of the solution. Eh and pH are continuously measured by hydrothermal Cormet© electrodes implanted in a titanium hydrothermal reactor. The transformation, or not, of the solid fraction was examined post-mortem. These experiments evidenced the role of Al colloids as precursor of cements composed of kaolinite and boehmite, and the passivation of amorphous silica (becoming unreactive) likely by sorption of aqueous iron. But no Fe-bearing was formed by contrast to many published studies on the Fe-clay interactions in the nuclear waste storage, and by contrast with basic thermodynamic predictions. Conclusion: The Fe-clays and steam generators contexts imply relatively close aqueous environments: hydrothermal, reduced, diluted, neutral to slightly alkaline. The main difference is the status of iron: ferric/ferrous (magnetite) in the steam generators, metallic in the Fe-clay experiments. The concentration of aqueous iron when supplied by magnetite is low and does not allow its incorporation in secondary phases. By contrast, aqueous ferrous iron released by the corrosion of steel is not limited by the source, rather by the sink, and produces Fe-rich silicates. This example illustrates the discrepancy between complex mineral reactions and oversimplified predictions when sorption/passivation and nucleation/growth constraints are ignored. Reference: [1] Lanson et al. (2012) Amer. Min. 97, 864-871. [2] Lantenois et al. (2005) Clays & Clay Min. 53, 597-612. [3] Mosser-Ruck et al. (2010) Clays & Clay Min. 58, 280-291. [4] Perronnet et al. (2008) App. Clay Sci. 38, 187-202. [5] Osacky et al. (2010) App. Clay Sci. 50, 237-244. [6] Guillaume et al. (2003) Clay Min. 38, 281-302. [7] Rivard et al. (2013) Amer. Mineral. 98, 163-180. [8] Svensson and Hansen (2013) Clays & Clay Min. 61, 566-579.

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE PAGES

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.; ...

2017-03-11

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

Crack growth testing on Cold Worked Alloy 690 in Primary Water Environment

NASA Astrophysics Data System (ADS)

Tice, David R.; Medway, Stuart L.; Platts, Norman; Stairmand, John W.

While plant experience so far has shown excellent resistance of Alloy 690 to stress corrosion cracking in PWR primary water environments, laboratory tests have reported that susceptibility may be enhanced substantially by non-uniform cold working, particularly when the plane of crack growth is in the plane of rolling or forging. The Alloy 690 program aims to further the understanding of the mechanisms behind this susceptibility and the heat-to-heat variability reported for different materials.

Advanced Information Systems Design: Technical Basis and Human Factors Review Guidance

DTIC Science & Technology

2000-03-01

D ., Wise, J ., and Hanes, L., "An Evaluation of Nuclear Power Plant Safety Parameter Display Systems," Proceedings of the Human Factors Society 25th...Reactor (PWR) (Source: Reprinted with permission from Woods, D ., Wise, J ., and Hanes, L., "An Evaluation of Nuclear Power Plant Safety Parameter...Dials display rpCJni?3 (b) Fluid-Tanks display B (c) Seesaw display I 72 CF \\^- J B ’ V ’II ’ ( d ) Mimic display B E * • \\ ^r 7

Hot zero power reactor calculations using the Insilico code

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; ...

2016-03-18

In this paper we describe the reactor physics simulation capabilities of the insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that the insilico SP N solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various PWR problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

33 CFR 208.11 - Regulations for use of storage allocated for flood control or navigation and/or project operation...

Code of Federal Regulations, 2011 CFR

2011-07-01

... VA Pwr. Glen Elder Dam & Waconda Lk KS Mitchel Solomon R FIM 722.3204.8 1488.31455.6 1455.61428.0... 820 PL 78-534 USBR. Kirwin Dam & Res KS Phillips N Fork Solomon R F ICR 215.1 89.6 1757.3 1729.2 1729... Webster Dam & Res KS Rocks S Fork Solomon R F IRC 183.4 72.1 1923.7 1892.5 1892.5 1860.0 8480 3772 3772...

33 CFR 208.11 - Regulations for use of storage allocated for flood control or navigation and/or project operation...

Code of Federal Regulations, 2010 CFR

2010-07-01

... VA Pwr. Glen Elder Dam & Waconda Lk KS Mitchel Solomon R FIM 722.3204.8 1488.31455.6 1455.61428.0... 820 PL 78-534 USBR. Kirwin Dam & Res KS Phillips N Fork Solomon R F ICR 215.1 89.6 1757.3 1729.2 1729... Webster Dam & Res KS Rocks S Fork Solomon R F IRC 183.4 72.1 1923.7 1892.5 1892.5 1860.0 8480 3772 3772...

Determination of tube-to-tube support interaction characteristics. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haslinger, K.H.

Tube-to-tube support interaction characteristics were determined on a multi-span tube geometry representative of the hot-leg side of the C-E, System 80 steam generator design. Results will become input for an autoclave type wear test program on steam generator tubes, performed by Kraftwerk Union (KWU). Correlation of test data reported here with similar data obtained from the wear tests will be performed in an attempt to make predictions about the long-term fretting behavior of steam generator tubes.

Advanced techniques for repair of irradiated PWR fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Westphal, M.

Kraftwerk union has recently designed and built a portable repair unit for use in nuclear power plants for repair of defective fuel assemblies where space limitations do not allow permanent installation of repair equipment. This new equipment is designed to be easily disassembled and decontaminated. The main component of the equipment is the fuel assembly reconstitution unit (FARU) which is placed on the floor of the spent fuel pool. The use of the FARU is described in the paper.

Survey and Analysis of Environmental Requirements for Shipboard Electronic Equipment Applications. Appendix A. Volume 2.

DTIC Science & Technology

1991-07-31

INTELLIGENT SCSI DMV-719 MAS MIL CONTROLLER DY-4 SYSTEMS BYTE-WIDE MEMORY CARD DMV-536 MEM MIL DY-4 SYSTEMS POWER SUPPLY UNIT DMV-870 PWR MIL P age No. 5 06/10...FORCE COMPUTERS PROCESSOR CPU-386 SERIES SBC COM FORCE COMPUTERS ADVANCED SYSTEM CONTROL ASCU -1/2 SBC COM UNITI FORCE COMPUTERS GRAPHICS CONTROLLER AGC...RECORD VENDOR: JANZ COMPUTER AG DIVISION: VENDOR ADDRESS: Im Doerener Feld 3 D-4790 Paderborn Germany MARKETING: Johannes Kunz TECHNICAL: Arnulf

Flooding Experiments and Modeling for Improved Reactor Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Solmos, M.; Hogan, K. J.; Vierow, K.

2008-09-14

Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge linemore » can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schulz, H.

In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A =more » open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.« less

Development of modified MDA (M-MDA), PWR fuel cladding tube for high duty operation in future

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watanabe, Seiichi; Kido, Toshiya; Arakawa, Yasushi

2007-07-01

A new cladding material of M-MDA has been developed in order to prepare for a strong growing demand for advanced fuel which can maintain its integrity even under high duties due to more efficient operation such as higher burnup, higher LHR, and longer operation cycle which will contribute the suppression of environmental burdens like CO{sub 2} emission. The main aim of M-MDA is to have excellent corrosion resistance while the other properties are inherited from MDA, which has been adopted to the step 2 fuel, instead of Zry-4, of Japanese PWR plant whose upper limit of assembly discharged burnup ismore » 55 MWd/kgU. And we could confirm that the main aim of M-MDA was achieved by means of out-of-pile tests. In order to confirm improvement of corrosion resistance of M-MDA in the actual operation, irradiation test of M-MDA in the commercial reactor of Vandellos II is ongoing. The latest results of on-site examination after every end of cycle showed that oxide thickness of M-MDA-SR was much smaller than that of MDA at rod discharged burnup of approximately 60 MWd/kgU. The final irradiation cycle was completed on April 2007 and then we will obtain corrosion data of M-MDA over 70 MWd/kgU. M-MDA is a candidate alloy for advanced fuel under higher duty usage. (authors)« less

ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas

2007-07-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code systemmore » ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)« less

Development and Application of Laser Peening System for PWR Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masaki Yoda; Itaru Chida; Satoshi Okada

2006-07-01

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion crackingmore » (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simonen, F.A.; Johnson, K.I.; Liebetrau, A.M.

The VISA-II (Vessel Integrity Simulation Analysis code was originally developed as part of the NRC staff evaluation of pressurized thermal shock. VISA-II uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics methods are used to model crack initiation and propagation. Parameters for initial crack size and location, copper content, initial reference temperature of the nil-ductility transition, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents an upgraded version of themore » original VISA code as described in NUREG/CR-3384. Improvements include a treatment of cladding effects, a more general simulation of flaw size, shape and location, a simulation of inservice inspection, an updated simulation of the reference temperature of the nil-ductility transition, and treatment of vessels with multiple welds and initial flaws. The code has been extensively tested and verified and is written in FORTRAN for ease of installation on different computers. 38 refs., 25 figs.« less

Current and planned numerical development for improving computing performance for long duration and/or low pressure transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Faydide, B.

1997-07-01

This paper presents the current and planned numerical development for improving computing performance in case of Cathare applications needing real time, like simulator applications. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the general characteristics of the code are presented, dealing with physical models, numerical topics, and validation strategy. Then, the current and planned applications of Cathare in the field of simulators are discussed. Some of these applications were made in the past, using a simplified and fast-running version of Cathare (Cathare-Simu); the status of the numerical improvements obtained withmore » Cathare-Simu is presented. The planned developments concern mainly the Simulator Cathare Release (SCAR) project which deals with the use of the most recent version of Cathare inside simulators. In this frame, the numerical developments are related with the speed up of the calculation process, using parallel processing and improvement of code reliability on a large set of NPP transients.« less

Posttest RELAP5 simulations of the Semiscale S-UT series experiments. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leonard, M.T.

The RELAP5/MOD1 computer code was used to perform posttest calculations, simulating six experiments, run in the Semiscale Mod-2A facility, investigating the effects of upper head injection on small break transient behavior. The results of these calculations and corresponding test data are presented in this report. An evaluation is made of the capability of RELAP5 to calculate the thermal-hydraulic response of the Mod-2A system over a spectrum of break sizes, with and without the use of upper head injection.

Evaluation of Potential Locations for Siting Small Modular Reactors near Federal Energy Clusters to Support Federal Clean Energy Goals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy J.; Omitaomu, Olufemi A.

2014-09-01

Geographic information systems (GIS) technology was applied to analyze federal energy demand across the contiguous US. Several federal energy clusters were previously identified, including Hampton Roads, Virginia, which was subsequently studied in detail. This study provides an analysis of three additional diverse federal energy clusters. The analysis shows that there are potential sites in various federal energy clusters that could be evaluated further for placement of an integral pressurized-water reactor (iPWR) to support meeting federal clean energy goals.

Integrated Systems Performance Assessment for the Evaluation of Space Nuclear Reactor Design Concepts (Phase 1: Demonstration of the Methodology).

DTIC Science & Technology

1992-11-01

Incorporated. Each design is characterized by a moderated core, a NaK pumped loop primary coolant system, and a potassium heat pipe radiator as the...1 1 10 1 RelHX 1 2 10 2 nRel HX 3 3 RelSS nRelSS Irr 4 3 7 8 9 io 2 + 2 + 2 + 2 nRel Pwr nRel NaK nRel RC nRel HX 1 1 11 1 RelSS 1 2 11 2 nRel SS 3 3

Use of artificial intelligence in severe accident diagnosis for PWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wu, Zheng; Okrent, D.; Kastenberg, W.E.

1995-12-31

A combination approach of an expert system and neural networks is used to implement a prototype severe accident diagnostic system which would monitor the progression of the severe accident and provide necessary plant status information to assist the plant staff in accident management during the accident. The station blackout accident in a pressurized water reactor (PWR) is used as the study case. The current phase of research focus is on distinguishing different primary system failure modes and following the accident transient before and up to vessel breach.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Conceptual Designing of a Reduced Moderation Pressurized Water Reactor by Use of MVP and MVP-BURN

NASA Astrophysics Data System (ADS)

Kugo, T.

A conceptual design of a seed-blanket assembly PWR core with a complicated geometry and a strong heterogeneity has been carried forward by use of the continuous-energy Monte Carlo method. Through parametric survey calculations by repeated use of MVP and a lattice burn-up calculation by MVP-BURN, a seed-blanket assembly configuration suitable for a concept of RMWR has been established, by evaluating precisely reactivity, a conversion ratio and a coolant void reactivity coefficient in a realistic computation time on a super computer.

TREAT Neutronics Analysis and Design Support, Part II: Multi-SERTTA-CAL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Woolstenhulme, Nicolas E.; Hill, Connie M.

2016-08-01

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. In integral aspect of prior TREAT transientmore » testing was the incorporation of calibration experiments to experimentally evaluate and validate test conditions in preparation of the actual fuel testing. The calibration experiment package established the test parameter conditions to support fine-tuning of the computational models to deliver the required energy deposition to the fuel samples. The calibration vehicle was designed to be as near neutronically equivalent to the experiment vehicle as possible to minimize errors between the calibration and final tests. The Multi-SERTTA-CAL vehicle was designed to serve as the calibration vehicle supporting Multi-SERTTA experimentation. Models of the Multi-SERTTA-CAL vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries; these results were then compared against those performed for Multi-SERTTA to determine the similarity and possible design modification necessary prior to construction of these experiment vehicles. The estimated reactivity insertion worth into the TREAT core is very similar between the two vehicle designs, with the primary physical difference being a hollow Inconel tube running down the length of the calibration vehicle. Calculations of PCF indicate that on average there is a reduction of approximately 6.3 and 12.6%, respectively, for PWR fuel rodlets irradiated under wet and dry conditions. Changes to the primary or secondary vessel structure in the calibration vehicle can be performed to offset this discrepancy and maintain neutronic equivalency. Current possible modifications to the calibration vehicle include reduction of the primary vessel wall thickness, swapping Zircaloy-4 for stainless steel 316 in the secondary containment, or slight modification to the temperature and pressure of the water environment within the primary vessel. Removal of some of the instrumentation within the calibration vehicle can also serve to slightly increase the PCF. Future efforts include further modification and optimization of the Multi-SERTTA and Multi-SERTTA-CAL designs in preparation of actual TREAT transient testing. Experimental results from both test vehicles will be compared against calculational results and methods to provide validation and support additional neutronics analyses.« less

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernnat, W.; Buck, M.; Mattes, M.

The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code ormore » memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

CECE: A Deep Throttling Demonstrator Cryogenic Engine for NASA's Lunar Lander

NASA Technical Reports Server (NTRS)

Giuliano, Victor J.; Leonard, Timothy G.; Adamski, Walter M.; Kim, Tony S.

2007-01-01

As one of the first technology development programs awarded under NASA's Vision for Space Exploration, the Pratt & Whitney Rocketdyne (PWR) Deep Throttling, Common Extensible Cryogenic Engine (CECE) program was selected by NASA in November 2004 to begin technology development and demonstration toward a deep throttling, cryogenic Lunar Lander engine for use across multiple human and robotic lunar exploration mission segments with extensibility to Mars. The CECE program leverages the maturity and previous investment of a flight-proven hydrogen/oxygen expander cycle engine, the RL10, to develop and demonstrate an unprecedented combination of reliability, safety, durability, throttlability, and restart capabilities in a high-energy, cryogenic engine. NASA Marshall Space Flight Center and NASA Glenn Research Center personnel were integral design and analysis team members throughout the requirements assessment, propellant studies and the deep throttling demonstrator elements of the program. The testbed selected for the initial deep throttling demonstration phase of this program was a minimally modified RL10 engine, allowing for maximum current production engine commonality and extensibility with minimum program cost. In just nine months from technical program start, CECE Demonstrator No. 1 engine testing in April/May 2006 at PWR's E06 test stand successfully demonstrated in excess of 10:1 throttling of the hydrogen/oxygen expander cycle engine. This test provided an early demonstration of a viable, enabling cryogenic propulsion concept with invaluable system-level technology data acquisition toward design and development risk mitigation for both the subsequent CECE Demonstrator No. 2 program and to the future Lunar Lander Design, Development, Test and Evaluation effort.

EMERALD REV. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

1974-03-01

The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place inmore » the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies.« less

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

Calculating the Responses of Self-Powered Radiation Detectors.

NASA Astrophysics Data System (ADS)

Thornton, D. A.

Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual response mechanisms.

Airframe noise of a small model transport aircraft and scaling effects. [Boeing 747

NASA Technical Reports Server (NTRS)

Shearin, J. G.

1981-01-01

Airframe noise of a 0.01 scale model Boeing 747 wide-body transport was measured in the Langley Anechoic Noise Facility. The model geometry simulated the landing and cruise configurations. The model noise was found to be similar in noise characteristics to that possessed by a 0.03 scale model 747. The 0.01 scale model noise data scaled to within 3 dB of full scale data using the same scaling relationships as that used to scale the 0.03 scale model noise data. The model noise data are compared with full scale noise data, where the full scale data are calculated using the NASA aircraft noise prediction program.

Airframe noise of a small model transport aircraft and scaling effects

NASA Astrophysics Data System (ADS)

Shearin, J. G.

1981-05-01

Airframe noise of a 0.01 scale model Boeing 747 wide-body transport was measured in the Langley Anechoic Noise Facility. The model geometry simulated the landing and cruise configurations. The model noise was found to be similar in noise characteristics to that possessed by a 0.03 scale model 747. The 0.01 scale model noise data scaled to within 3 dB of full scale data using the same scaling relationships as that used to scale the 0.03 scale model noise data. The model noise data are compared with full scale noise data, where the full scale data are calculated using the NASA aircraft noise prediction program.

Methodology, status and plans for development and assessment of Cathare code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestion, D.; Barre, F.; Faydide, B.

1997-07-01

This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests ormore » integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.« less

Preliminary Investigation of Time Remaining Display on the Computer-based Emergency Operating Procedure

NASA Astrophysics Data System (ADS)

Suryono, T. J.; Gofuku, A.

2018-02-01

One of the important thing in the mitigation of accidents in nuclear power plant accidents is time management. The accidents should be resolved as soon as possible in order to prevent the core melting and the release of radioactive material to the environment. In this case, operators should follow the emergency operating procedure related with the accident, in step by step order and in allowable time. Nowadays, the advanced main control rooms are equipped with computer-based procedures (CBPs) which is make it easier for operators to do their tasks of monitoring and controlling the reactor. However, most of the CBPs do not include the time remaining display feature which informs operators of time available for them to execute procedure steps and warns them if the they reach the time limit. Furthermore, the feature will increase the awareness of operators about their current situation in the procedure. This paper investigates this issue. The simplified of emergency operating procedure (EOP) of steam generator tube rupture (SGTR) accident of PWR plant is applied. In addition, the sequence of actions on each step of the procedure is modelled using multilevel flow modelling (MFM) and influenced propagation rule. The prediction of action time on each step is acquired based on similar case accidents and the Support Vector Regression. The derived time will be processed and then displayed on a CBP user interface.

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

Approach to numerical safety guidelines based on a core melt criterion. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azarm, M.A.; Hall, R.E.

1982-01-01

A plausible approach is proposed for translating a single level criterion to a set of numerical guidelines. The criterion for core melt probability is used to set numerical guidelines for various core melt sequences, systems and component unavailabilities. These guidelines can be used as a means for making decisions regarding the necessity for replacing a component or improving part of a safety system. This approach is applied to estimate a set of numerical guidelines for various sequences of core melts that are analyzed in Reactor Safety Study for the Peach Bottom Nuclear Power Plant.

Spent fuel data base: commercial light water reactors. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Material control and accountancy at EDF PWR plants; GCN: Gestion du Combustible Nucleaire

DOE Office of Scientific and Technical Information (OSTI.GOV)

de Cormis, F.

1991-01-01

The paper describes the comprehensive system which is developed and implemented at Electricite de France to provide a single reliable nuclear material control and accounting system for all nuclear plants. This software aims at several objectives among which are: the control and the accountancy of nuclear material at the plant, the optimization of the consistency of data by minimizing the possibility of transcription errors, the fulfillment of the statutory requirements by automatic transfer of reports to national and international safeguards authorities, the servicing of other EDF users of nuclear material data for technical or commercial purposes.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

Assignment of boundary conditions in embedded ground water flow models

USGS Publications Warehouse

Leake, S.A.

1998-01-01

Many small-scale ground water models are too small to incorporate distant aquifer boundaries. If a larger-scale model exists for the area of interest, flow and head values can be specified for boundaries in the smaller-scale model using values from the larger-scale model. Flow components along rows and columns of a large-scale block-centered finite-difference model can be interpolated to compute horizontal flow across any segment of a perimeter of a small-scale model. Head at cell centers of the larger-scale model can be interpolated to compute head at points on a model perimeter. Simple linear interpolation is proposed for horizontal interpolation of horizontal-flow components. Bilinear interpolation is proposed for horizontal interpolation of head values. The methods of interpolation provided satisfactory boundary conditions in tests using models of hypothetical aquifers.Many small-scale ground water models are too small to incorporate distant aquifer boundaries. If a larger-scale model exists for the area of interest, flow and head values can be specified for boundaries in the smaller-scale model using values from the larger-scale model. Flow components along rows and columns of a large-scale block-centered finite-difference model can be interpolated to compute horizontal flow across any segment of a perimeter of a small-scale model. Head at cell centers of the larger.scale model can be interpolated to compute head at points on a model perimeter. Simple linear interpolation is proposed for horizontal interpolation of horizontal-flow components. Bilinear interpolation is proposed for horizontal interpolation of head values. The methods of interpolation provided satisfactory boundary conditions in tests using models of hypothetical aquifers.

Pore-scale simulation of microbial growth using a genome-scale metabolic model: Implications for Darcy-scale reactive transport

NASA Astrophysics Data System (ADS)

Tartakovsky, G. D.; Tartakovsky, A. M.; Scheibe, T. D.; Fang, Y.; Mahadevan, R.; Lovley, D. R.

2013-09-01

Recent advances in microbiology have enabled the quantitative simulation of microbial metabolism and growth based on genome-scale characterization of metabolic pathways and fluxes. We have incorporated a genome-scale metabolic model of the iron-reducing bacteria Geobacter sulfurreducens into a pore-scale simulation of microbial growth based on coupling of iron reduction to oxidation of a soluble electron donor (acetate). In our model, fluid flow and solute transport is governed by a combination of the Navier-Stokes and advection-diffusion-reaction equations. Microbial growth occurs only on the surface of soil grains where solid-phase mineral iron oxides are available. Mass fluxes of chemical species associated with microbial growth are described by the genome-scale microbial model, implemented using a constraint-based metabolic model, and provide the Robin-type boundary condition for the advection-diffusion equation at soil grain surfaces. Conventional models of microbially-mediated subsurface reactions use a lumped reaction model that does not consider individual microbial reaction pathways, and describe reactions rates using empirically-derived rate formulations such as the Monod-type kinetics. We have used our pore-scale model to explore the relationship between genome-scale metabolic models and Monod-type formulations, and to assess the manifestation of pore-scale variability (microenvironments) in terms of apparent Darcy-scale microbial reaction rates. The genome-scale model predicted lower biomass yield, and different stoichiometry for iron consumption, in comparison to prior Monod formulations based on energetics considerations. We were able to fit an equivalent Monod model, by modifying the reaction stoichiometry and biomass yield coefficient, that could effectively match results of the genome-scale simulation of microbial behaviors under excess nutrient conditions, but predictions of the fitted Monod model deviated from those of the genome-scale model under conditions in which one or more nutrients were limiting. The fitted Monod kinetic model was also applied at the Darcy scale; that is, to simulate average reaction processes at the scale of the entire pore-scale model domain. As we expected, even under excess nutrient conditions for which the Monod and genome-scale models predicted equal reaction rates at the pore scale, the Monod model over-predicted the rates of biomass growth and iron and acetate utilization when applied at the Darcy scale. This discrepancy is caused by an inherent assumption of perfect mixing over the Darcy-scale domain, which is clearly violated in the pore-scale models. These results help to explain the need to modify the flux constraint parameters in order to match observations in previous applications of the genome-scale model at larger scales. These results also motivate further investigation of quantitative multi-scale relationships between fundamental behavior at the pore scale (where genome-scale models are appropriately applied) and observed behavior at larger scales (where predictions of reactive transport phenomena are needed).

Pore-scale simulation of microbial growth using a genome-scale metabolic model: Implications for Darcy-scale reactive transport

NASA Astrophysics Data System (ADS)

Scheibe, T. D.; Tartakovsky, G.; Tartakovsky, A. M.; Fang, Y.; Mahadevan, R.; Lovley, D. R.

2012-12-01

Recent advances in microbiology have enabled the quantitative simulation of microbial metabolism and growth based on genome-scale characterization of metabolic pathways and fluxes. We have incorporated a genome-scale metabolic model of the iron-reducing bacteria Geobacter sulfurreducens into a pore-scale simulation of microbial growth based on coupling of iron reduction to oxidation of a soluble electron donor (acetate). In our model, fluid flow and solute transport is governed by a combination of the Navier-Stokes and advection-diffusion-reaction equations. Microbial growth occurs only on the surface of soil grains where solid-phase mineral iron oxides are available. Mass fluxes of chemical species associated with microbial growth are described by the genome-scale microbial model, implemented using a constraint-based metabolic model, and provide the Robin-type boundary condition for the advection-diffusion equation at soil grain surfaces. Conventional models of microbially-mediated subsurface reactions use a lumped reaction model that does not consider individual microbial reaction pathways, and describe reactions rates using empirically-derived rate formulations such as the Monod-type kinetics. We have used our pore-scale model to explore the relationship between genome-scale metabolic models and Monod-type formulations, and to assess the manifestation of pore-scale variability (microenvironments) in terms of apparent Darcy-scale microbial reaction rates. The genome-scale model predicted lower biomass yield, and different stoichiometry for iron consumption, in comparison to prior Monod formulations based on energetics considerations. We were able to fit an equivalent Monod model, by modifying the reaction stoichiometry and biomass yield coefficient, that could effectively match results of the genome-scale simulation of microbial behaviors under excess nutrient conditions, but predictions of the fitted Monod model deviated from those of the genome-scale model under conditions in which one or more nutrients were limiting. The fitted Monod kinetic model was also applied at the Darcy scale; that is, to simulate average reaction processes at the scale of the entire pore-scale model domain. As we expected, even under excess nutrient conditions for which the Monod and genome-scale models predicted equal reaction rates at the pore scale, the Monod model over-predicted the rates of biomass growth and iron and acetate utilization when applied at the Darcy scale. This discrepancy is caused by an inherent assumption of perfect mixing over the Darcy-scale domain, which is clearly violated in the pore-scale models. These results help to explain the need to modify the flux constraint parameters in order to match observations in previous applications of the genome-scale model at larger scales. These results also motivate further investigation of quantitative multi-scale relationships between fundamental behavior at the pore scale (where genome-scale models are appropriately applied) and observed behavior at larger scales (where predictions of reactive transport phenomena are needed).

Pore-scale simulation of microbial growth using a genome-scale metabolic model: Implications for Darcy-scale reactive transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tartakovsky, Guzel D.; Tartakovsky, Alexandre M.; Scheibe, Timothy D.

2013-09-07

Recent advances in microbiology have enabled the quantitative simulation of microbial metabolism and growth based on genome-scale characterization of metabolic pathways and fluxes. We have incorporated a genome-scale metabolic model of the iron-reducing bacteria Geobacter sulfurreducens into a pore-scale simulation of microbial growth based on coupling of iron reduction to oxidation of a soluble electron donor (acetate). In our model, fluid flow and solute transport is governed by a combination of the Navier-Stokes and advection-diffusion-reaction equations. Microbial growth occurs only on the surface of soil grains where solid-phase mineral iron oxides are available. Mass fluxes of chemical species associated withmore » microbial growth are described by the genome-scale microbial model, implemented using a constraint-based metabolic model, and provide the Robin-type boundary condition for the advection-diffusion equation at soil grain surfaces. Conventional models of microbially-mediated subsurface reactions use a lumped reaction model that does not consider individual microbial reaction pathways, and describe reactions rates using empirically-derived rate formulations such as the Monod-type kinetics. We have used our pore-scale model to explore the relationship between genome-scale metabolic models and Monod-type formulations, and to assess the manifestation of pore-scale variability (microenvironments) in terms of apparent Darcy-scale microbial reaction rates. The genome-scale model predicted lower biomass yield, and different stoichiometry for iron consumption, in comparisonto prior Monod formulations based on energetics considerations. We were able to fit an equivalent Monod model, by modifying the reaction stoichiometry and biomass yield coefficient, that could effectively match results of the genome-scale simulation of microbial behaviors under excess nutrient conditions, but predictions of the fitted Monod model deviated from those of the genome-scale model under conditions in which one or more nutrients were limiting. The fitted Monod kinetic model was also applied at the Darcy scale; that is, to simulate average reaction processes at the scale of the entire pore-scale model domain. As we expected, even under excess nutrient conditions for which the Monod and genome-scale models predicted equal reaction rates at the pore scale, the Monod model over-predicted the rates of biomass growth and iron and acetate utilization when applied at the Darcy scale. This discrepancy is caused by an inherent assumption of perfect mixing over the Darcy-scale domain, which is clearly violated in the pore-scale models. These results help to explain the need to modify the flux constraint parameters in order to match observations in previous applications of the genome-scale model at larger scales. These results also motivate further investigation of quantitative multi-scale relationships between fundamental behavior at the pore scale (where genome-scale models are appropriately applied) and observed behavior at larger scales (where predictions of reactive transport phenomena are needed).« less

Strategies for efficient numerical implementation of hybrid multi-scale agent-based models to describe biological systems

PubMed Central

Cilfone, Nicholas A.; Kirschner, Denise E.; Linderman, Jennifer J.

2015-01-01

Biologically related processes operate across multiple spatiotemporal scales. For computational modeling methodologies to mimic this biological complexity, individual scale models must be linked in ways that allow for dynamic exchange of information across scales. A powerful methodology is to combine a discrete modeling approach, agent-based models (ABMs), with continuum models to form hybrid models. Hybrid multi-scale ABMs have been used to simulate emergent responses of biological systems. Here, we review two aspects of hybrid multi-scale ABMs: linking individual scale models and efficiently solving the resulting model. We discuss the computational choices associated with aspects of linking individual scale models while simultaneously maintaining model tractability. We demonstrate implementations of existing numerical methods in the context of hybrid multi-scale ABMs. Using an example model describing Mycobacterium tuberculosis infection, we show relative computational speeds of various combinations of numerical methods. Efficient linking and solution of hybrid multi-scale ABMs is key to model portability, modularity, and their use in understanding biological phenomena at a systems level. PMID:26366228

The effects of heat treatment on the chromium depletion, precipitate evolution, and corrosion resistance of INCONEL alloy 690

NASA Astrophysics Data System (ADS)

Kai, J. J.; Yu, G. P.; Tsai, C. H.; Liu, M. N.; Yao, S. C.

1989-10-01

A series of heat treatments were performed to study the sensitization and the stress corrosion cracking (SCC) behavior of INCONEL Alloy 690. The microstructural evaluation and the chromium depletion near grain boundaries were carefully studied using analytical electron microscopy (AEM). The measured chromium depletion profiles were matched well to the calculated results from a thermodynamic/kinetic model. The constant extension rate test (CERT) was performed in the solution containing 0.001 M sodium thiosulfate (Na2S2O3) to study the SCC resistance of this alloy. The Huey test was also performed in a boiling 65 pct HNO3 solution for 48 hours to study the intergranular attack (IGA) resistance of this alloy. Both tests showed that INCONEL 690 has very good corrosion resistance. It is believed that the superior IGA and SCC resistances of this alloy are due to the high chromium concentration (≈30 wt pct). It is concluded in this study that INCONEL 690 may be a better alloy than INCONEL 600 for use as the steam generator (S/G) tubing material for pressurized water reactors (PWR's)

CASL L1 Milestone report : CASL.P4.01, sensitivity and uncertainty analysis for CIPS with VIPRE-W and BOA.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sung, Yixing; Adams, Brian M.; Secker, Jeffrey R.

2011-12-01

The CASL Level 1 Milestone CASL.P4.01, successfully completed in December 2011, aimed to 'conduct, using methodologies integrated into VERA, a detailed sensitivity analysis and uncertainty quantification of a crud-relevant problem with baseline VERA capabilities (ANC/VIPRE-W/BOA).' The VUQ focus area led this effort, in partnership with AMA, and with support from VRI. DAKOTA was coupled to existing VIPRE-W thermal-hydraulics and BOA crud/boron deposit simulations representing a pressurized water reactor (PWR) that previously experienced crud-induced power shift (CIPS). This work supports understanding of CIPS by exploring the sensitivity and uncertainty in BOA outputs with respect to uncertain operating and model parameters. Thismore » report summarizes work coupling the software tools, characterizing uncertainties, and analyzing the results of iterative sensitivity and uncertainty studies. These studies focused on sensitivity and uncertainty of CIPS indicators calculated by the current version of the BOA code used in the industry. Challenges with this kind of analysis are identified to inform follow-on research goals and VERA development targeting crud-related challenge problems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerard, R.; Malekian, C.; Meessen, O.

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports inmore » the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.« less

Isoniazid interaction with phosphatidylcholine-based membranes

NASA Astrophysics Data System (ADS)

Marques, Amanda Vicente; Marengo Trindade, Paulo; Marques, Sheylla; Brum, Tainá; Harte, Etienne; Rodrigues, Marieli Oliveira; D'Oca, Marcelo Gonçalves Montes; da Silva, Pedro Almeida; Pohlmann, Adriana R.; Alves, Isabel Dantas; de Lima, Vânia Rodrigues

2013-11-01

Interaction between the anti-tuberculosis drug isoniazid (INH) and phosphatidylcholine membranes was investigated in terms of: (i) drug affinity to a lipid bilayer and (ii) drug-induced changes in the dynamic properties of liposomes, such as membrane hydration state, polar head and non-polar acyl chain order and lipid phase transition behavior. These parameters were studied by plasmon waveguide resonance spectroscopy (PWR), UV-visible, horizontal attenuated total reflectance-Fourier transform infrared (HATR-FTIR), nuclear magnetic resonance (NMR) and differential scanning calorimetry (DSC) techniques. PWR measurements showed an INH membrane dissociation constant value of 0.031 μM to phosphatidylcholine bilayers. INH induced higher membrane perturbation in the plane which is perpendicular to the membrane plane. The INH saturation concentration in phosphatidylcholine liposomes was 170 μM. At this concentration, HATR-FTIR and NMR findings showed that INH may interact with the lipid polar head, increasing the number of hydrogen bonds in the phosphate region and enhancing the choline motional freedom. DSC measurements showed that, at 115 μM, INH was responsible for a decrease in lipid phase transition temperature of approximately 2 °C and had no influence in the lipid enthalpy variation (ΔH). However, at 170 μM, INH induced the reduction of the ΔH by approximately 52%, suggesting that the drug may increase the distance among lipid molecules and enhance the freedom of the lipid acyl chains methylene groups. This paper provides information on the effects of INH on membrane dynamics which is important to understand liposome targeting of the drug and for the development of anti-TB pharmacologic systems that not only are less susceptible to resistance but also have low toxicity.

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now.more » The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)« less

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

NASA Astrophysics Data System (ADS)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

J-2X Fuel Turbopump Point of Departure: The Performance of the J-2s Fuel Turbopump Inducer

NASA Technical Reports Server (NTRS)

Sargent, S. R.; Becht, D. G.; Mulder, A. D.

2008-01-01

To aid the J-2X program design effort with detailed performance and environment information, the J-2S fuel turbopump (FTP) inducer has undergone a thorough test series in both water and hydrogen. The test series utilizes both inducer only and a complete pump configuration to assess the inducer interaction to the overall turbopump system. The test goals include verification of suction performance against heritage J-2S data, head production, effects of thermodynamic suppression head (TSH), and evaluation of the inducer dynamic pressure caused by cavitation instabilities. Test facilities at both Pratt & Whitney Rocketdyne (PWR) and NASA s Stennis Space Center (SSC) are employed for the testing. The inducer only water test effort conducted at PWR established performance curves for suction performance, head production, and efficiency over a wide operating range. Because the heritage J-2S suction performance data set is in hydrogen, it is desired to obtain current suction performance data in hydrogen as well, thus avoiding the reliance on a theoretical TSH correlation for direct comparison. This data then provides an empirically based TSH correlation allowing for the comparison of water test suction data to system suction requirements. The FTP testing performed at SSC provides these suction performance relationships as well as inlet duct dynamic pressures during liquid hydrogen operation. The test effort successfully confirms the heritage J-2S suction performance and establishes the amount of TSH between water and hydrogen operation at the design flow coefficient. Correlating data is also obtained for cavitating instability frequency content, illustrating the validity of using the wide flow range water test data to predict hydrogen performance.

Corrosion behavior and oxide properties of Zr 1.1 wt%Nb 0.05 wt%Cu alloy

NASA Astrophysics Data System (ADS)

Park, Jeong-Yong; Choi, Byung-Kwon; Yoo, Seung Jo; Jeong, Yong Hwan

2006-12-01

The corrosion behavior and oxide properties of Zr-1.1 wt%Nb-0.05 wt%Cu (ZrNbCu) and Zircaloy-4 have been investigated. The corrosion rate of the ZrNbCu alloy was much lower than that of the Zirclaoy-4 in the 360 °C water and 360 °C PWR-simulating loop condition without a neutron flux and it was increased with an increase of the final annealing temperature from 470 °C to 570 °C. TEM observations revealed that the precipitates in the ZrNbCu were β-Nb and ZrNbFe-precipitate with β-Nb being more frequently observed and that the precipitates were more finely distributed in the ZrNbCu alloy. It was also observed that the oxides of the ZrNbCu and Zircaloy-4 consisted of two and seven layers, respectively, after 1000 days in the PWR-simulating loop condition and that the thickness of a fully-developed layer was higher in the ZrNbCu than in the Zircaloy-4. It was also found that the β-Nb in ZrNbCu was oxidized more slowly when compared to the Zr(Fe, Cr) 2 in Zirclaoy-4 when the precipitates in the oxide were observed by TEM. Cracks were observed in the vicinity of the oxidized Zr(Fe, Cr) 2, while no cracks were formed near β-Nb which had retained a metallic state. From the results obtained, it is suggested that the oxide formed on the ZrNbCu has a more protective nature against a corrosion when compared to that of the Zircaloy-4.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kruska, Karen; Zhai, Ziqing; Bruemmer, Stephen M.

Due to its superior resistance to corrosion and stress corrosion cracking (SCC), high Cr, Ni-base Alloy 690 is now commonly used in pressurized water reactors (PWRs). Even though highly cold-worked (CW) Alloy 690 has been shown to be susceptible to SCC crack growth in PWR primary water environments, an open question remains whether SCC initiation was possible for these materials under constant load test conditions. Testing has been performed on a series of CW alloy 690 CRDM tubing specimens at constant load for up to 9,220 hours in 360°C simulated PWR primary water. A companion paper will discuss the overallmore » testing approach and describe results on different alloy 690 heats and cold work levels. The focus of the current paper is to illustrate the use of focused ion beam (FIB), scanning electron microscopy (SEM) and transmission electron microscopy (TEM) for the high-resolution investigation of precursor damage and intergranular (IG) crack nucleation in these specimens. Three-dimensional (3D) FIB/SEM imaging has been conducted on a series of grain boundary (GB) damage precursors, such as IG small cavities, local corrosion and even shallow cracks observed at the specimen surface. Contrast variations and EDS mapping were used to distinguish oxides, carbides and cavities from the matrix material. Nanometer-sized cavities were observed associated with GB carbides in the highly CW specimens. Shallow IG cracks were present in the 30%CW specimens and exhibited oxidized crack flanks and a higher density of cavities ahead of the oxide front in all cases. The shape and distribution of carbides and cavities in the plane of the cracked GBs was analyzed in 3D to gain a mechanistic understanding of the processes that may be leading to crack initiation in highly CW alloy 690.« less

Rubber particle proteins, HbREF and HbSRPP, show different interactions with model membranes.

PubMed

Berthelot, Karine; Lecomte, Sophie; Estevez, Yannick; Zhendre, Vanessa; Henry, Sarah; Thévenot, Julie; Dufourc, Erick J; Alves, Isabel D; Peruch, Frédéric

2014-01-01

The biomembrane surrounding rubber particles from the hevea latex is well known for its content of numerous allergen proteins. HbREF (Hevb1) and HbSRPP (Hevb3) are major components, linked on rubber particles, and they have been shown to be involved in rubber synthesis or quality (mass regulation), but their exact function is still to be determined. In this study we highlighted the different modes of interactions of both recombinant proteins with various membrane models (lipid monolayers, liposomes or supported bilayers, and multilamellar vesicles) to mimic the latex particle membrane. We combined various biophysical methods (polarization-modulation-infrared reflection-adsorption spectroscopy (PM-IRRAS)/ellipsometry, attenuated-total reflectance Fourier-transform infrared (ATR-FTIR), solid-state nuclear magnetic resonance (NMR), plasmon waveguide resonance (PWR), fluorescence spectroscopy) to elucidate their interactions. Small rubber particle protein (SRPP) shows less affinity than rubber elongation factor (REF) for the membranes but displays a kind of "covering" effect on the lipid headgroups without disturbing the membrane integrity. Its structure is conserved in the presence of lipids. Contrarily, REF demonstrates higher membrane affinity with changes in its aggregation properties, the amyloid nature of REF, which we previously reported, is not favored in the presence of lipids. REF binds and inserts into membranes. The membrane integrity is highly perturbed, and we suspect that REF is even able to remove lipids from the membrane leading to the formation of mixed micelles. These two homologous proteins show affinity to all membrane models tested but neatly differ in their interacting features. This could imply differential roles on the surface of rubber particles. © 2013.

Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

NASA Astrophysics Data System (ADS)

Cook, William Gordon

Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

Downscaling modelling system for multi-scale air quality forecasting

NASA Astrophysics Data System (ADS)

Nuterman, R.; Baklanov, A.; Mahura, A.; Amstrup, B.; Weismann, J.

2010-09-01

Urban modelling for real meteorological situations, in general, considers only a small part of the urban area in a micro-meteorological model, and urban heterogeneities outside a modelling domain affect micro-scale processes. Therefore, it is important to build a chain of models of different scales with nesting of higher resolution models into larger scale lower resolution models. Usually, the up-scaled city- or meso-scale models consider parameterisations of urban effects or statistical descriptions of the urban morphology, whereas the micro-scale (street canyon) models are obstacle-resolved and they consider a detailed geometry of the buildings and the urban canopy. The developed system consists of the meso-, urban- and street-scale models. First, it is the Numerical Weather Prediction (HIgh Resolution Limited Area Model) model combined with Atmospheric Chemistry Transport (the Comprehensive Air quality Model with extensions) model. Several levels of urban parameterisation are considered. They are chosen depending on selected scales and resolutions. For regional scale, the urban parameterisation is based on the roughness and flux corrections approach; for urban scale - building effects parameterisation. Modern methods of computational fluid dynamics allow solving environmental problems connected with atmospheric transport of pollutants within urban canopy in a presence of penetrable (vegetation) and impenetrable (buildings) obstacles. For local- and micro-scales nesting the Micro-scale Model for Urban Environment is applied. This is a comprehensive obstacle-resolved urban wind-flow and dispersion model based on the Reynolds averaged Navier-Stokes approach and several turbulent closures, i.e. k -ɛ linear eddy-viscosity model, k - ɛ non-linear eddy-viscosity model and Reynolds stress model. Boundary and initial conditions for the micro-scale model are used from the up-scaled models with corresponding interpolation conserving the mass. For the boundaries a kind of Dirichlet condition is chosen to provide the values based on interpolation from the coarse to the fine grid. When the roughness approach is changed to the obstacle-resolved one in the nested model, the interpolation procedure will increase the computational time (due to additional iterations) for meteorological/ chemical fields inside the urban sub-layer. In such situations, as a possible alternative, the perturbation approach can be applied. Here, the effects of main meteorological variables and chemical species are considered as a sum of two components: background (large-scale) values, described by the coarse-resolution model, and perturbations (micro-scale) features, obtained from the nested fine resolution model.

Small-Scale Tests of MX Vertical Shelter Structures.

DTIC Science & Technology

1983-06-29

models were built with as much geometric and material similitude as practical. They 7were not identical to the 1/3-scale structures tested in the VST ...comparison with the 1/30-scale models and the 1/6-scale models, the 1/3-scale VST 7 models had different geometry (wall thickness variations), different...1/30-scale and 1/6-scale results with the 1/3-scale VST results. For example, the strains measured in the 1/3-scale ’B’ structure are about twice as

Static Aeroelastic Scaling and Analysis of a Sub-Scale Flexible Wing Wind Tunnel Model

NASA Technical Reports Server (NTRS)

Ting, Eric; Lebofsky, Sonia; Nguyen, Nhan; Trinh, Khanh

2014-01-01

This paper presents an approach to the development of a scaled wind tunnel model for static aeroelastic similarity with a full-scale wing model. The full-scale aircraft model is based on the NASA Generic Transport Model (GTM) with flexible wing structures referred to as the Elastically Shaped Aircraft Concept (ESAC). The baseline stiffness of the ESAC wing represents a conventionally stiff wing model. Static aeroelastic scaling is conducted on the stiff wing configuration to develop the wind tunnel model, but additional tailoring is also conducted such that the wind tunnel model achieves a 10% wing tip deflection at the wind tunnel test condition. An aeroelastic scaling procedure and analysis is conducted, and a sub-scale flexible wind tunnel model based on the full-scale's undeformed jig-shape is developed. Optimization of the flexible wind tunnel model's undeflected twist along the span, or pre-twist or wash-out, is then conducted for the design test condition. The resulting wind tunnel model is an aeroelastic model designed for the wind tunnel test condition.

Implementation of the SPH Procedure Within the MOOSE Finite Element Framework

NASA Astrophysics Data System (ADS)

Laurier, Alexandre

The goal of this thesis was to implement the SPH homogenization procedure within the MOOSE finite element framework at INL. Before this project, INL relied on DRAGON to do their SPH homogenization which was not flexible enough for their needs. As such, the SPH procedure was implemented for the neutron diffusion equation with the traditional, Selengut and true Selengut normalizations. Another aspect of this research was to derive the SPH corrected neutron transport equations and implement them in the same framework. Following in the footsteps of other articles, this feature was implemented and tested successfully with both the PN and S N transport calculation schemes. Although the results obtained for the power distribution in PWR assemblies show no advantages over the use of the SPH diffusion equation, we believe the inclusion of this transport correction will allow for better results in cases where either P N or SN are required. An additional aspect of this research was the implementation of a novel way of solving the non-linear SPH problem. Traditionally, this was done through a Picard, fixed-point iterative process whereas the new implementation relies on MOOSE's Preconditioned Jacobian-Free Newton Krylov (PJFNK) method to allow for a direct solution to the non-linear problem. This novel implementation showed a decrease in calculation time by a factor reaching 50 and generated SPH factors that correspond to those obtained through a fixed-point iterative process with a very tight convergence criteria: epsilon < 10-8. The use of the PJFNK SPH procedure also allows to reach convergence in problems containing important reflector regions and void boundary conditions, something that the traditional SPH method has never been able to achieve. At times when the PJFNK method cannot reach convergence to the SPH problem, a hybrid method is used where by the traditional SPH iteration forces the initial condition to be within the radius of convergence of the Newton method. This new method was tested on a simplified model of INL's TREAT reactor, a problem that includes very important graphite reflector regions as well as vacuum boundary conditions with great success. To demonstrate the power of PJFNK SPH on a more common case, the correction was applied to a simplified PWR reactor core from the BEAVRS benchmark that included 15 assemblies and the water reflector to obtain very good results. This opens up the possibility to apply the SPH correction to full reactor cores in order to reduce homogenization errors for use in transient or multi-physics calculations.

TREAT Neutronics Analysis and Design Support, Part I: Multi-SERTTA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Woolstenhulme, Nicolas E.; Hill, Connie M.

2016-08-01

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis ofmore » a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.74 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies.« less

Posttest TRAC-PD2/MOD1 predictions for FLECHT SEASET test 31504. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Booker, C.P.

TRAC-PD2/MOD1 is a publicly released version of TRAC that is used primarily to analyze large-break loss-of-coolant accidents in pressurized-water reactors (PWRs). TRAC-PD2 can calculate, among other things, reflood phenomena. TRAC posttest predictions are compared with test 31504 reflood data from the Full-Length Emergency Core Heat Transfer (FLECHT) System Effects and Separate Effects Tests (SEASET) facility. A false top-down quench is predicted near the top of the core and the subcooling is underpredicted at the bottom of the core. However, the overall TRAC predictions are good, especially near the center of the core.

Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

DOEpatents

DeVolpi, A.

1984-07-20

The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

Stress corrosion crack initiation of alloy 600 in PWR primary water

DOE PAGES

Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.; ...

2017-04-27

Stress corrosion crack (SCC) initiation of three mill-annealed alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and intergranular SCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Lastly, we discuss processes controlling the SCC initiation in mill-annealed alloy 600.

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Clarno, Kevin T.; Gentry, Cole

2017-03-01

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

Instability study for LOFT for L2-1, L2-2, and L2-3 pretest steady-state operating conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eide, S.A.

The results are presented of a thermal-hydrodynamic flow instability study of the LOFT reactor for the L2-1, L2-2, and L2-3 pretest steady-state operating conditions. Comparison is made between the LOFT reactor and a typical PWR, and the effects on stability of differences in operating parameters and geometry are discussed. Results indicate that the LOFT reactor will be thermal-hydrodynamically stable for nominal and worst case operating conditions. The study supports the LOFT Experimental Safety Analyses for the L2-1, L2-2, and L2-3 tests.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Stress corrosion crack initiation of alloy 600 in PWR primary water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.

Stress corrosion crack (SCC) initiation of three mill-annealed alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and intergranular SCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Lastly, we discuss processes controlling the SCC initiation in mill-annealed alloy 600.

Stress corrosion of low alloy steels used in external bolting on pressurised water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skeldon, P.; Hurst, P.; Smart, N.R.

1992-12-31

The stress corrosion cracking (SCC) susceptibility of AISI 4140 and AISI 4340 steels has been evaluated in five environments, three simulating a leaking aqueous boric acid environment and two simulating ambient external conditions ie moist air and salt spray. Both steels were found to be highly susceptible to SCC in all environments at hardnesses of 400 VPN and above. The susceptibility was greatly reduced at hardnesses below 330 VPN but in one environment, viz refluxing PWR primary water, SCC was observed at hardnesses as low as 260VPN. Threshold stress intensities for SCC were frequently lower than those in the literature.

Scale effect challenges in urban hydrology highlighted with a distributed hydrological model

NASA Astrophysics Data System (ADS)

Ichiba, Abdellah; Gires, Auguste; Tchiguirinskaia, Ioulia; Schertzer, Daniel; Bompard, Philippe; Ten Veldhuis, Marie-Claire

2018-01-01

Hydrological models are extensively used in urban water management, development and evaluation of future scenarios and research activities. There is a growing interest in the development of fully distributed and grid-based models. However, some complex questions related to scale effects are not yet fully understood and still remain open issues in urban hydrology. In this paper we propose a two-step investigation framework to illustrate the extent of scale effects in urban hydrology. First, fractal tools are used to highlight the scale dependence observed within distributed data input into urban hydrological models. Then an intensive multi-scale modelling work is carried out to understand scale effects on hydrological model performance. Investigations are conducted using a fully distributed and physically based model, Multi-Hydro, developed at Ecole des Ponts ParisTech. The model is implemented at 17 spatial resolutions ranging from 100 to 5 m. Results clearly exhibit scale effect challenges in urban hydrology modelling. The applicability of fractal concepts highlights the scale dependence observed within distributed data. Patterns of geophysical data change when the size of the observation pixel changes. The multi-scale modelling investigation confirms scale effects on hydrological model performance. Results are analysed over three ranges of scales identified in the fractal analysis and confirmed through modelling. This work also discusses some remaining issues in urban hydrology modelling related to the availability of high-quality data at high resolutions, and model numerical instabilities as well as the computation time requirements. The main findings of this paper enable a replacement of traditional methods of model calibration by innovative methods of model resolution alteration based on the spatial data variability and scaling of flows in urban hydrology.

A Method for Estimating Noise from Full-Scale Distributed Exhaust Nozzles

NASA Technical Reports Server (NTRS)

Kinzie, Kevin W.; Schein, David B.

2004-01-01

A method to estimate the full-scale noise suppression from a scale model distributed exhaust nozzle (DEN) is presented. For a conventional scale model exhaust nozzle, Strouhal number scaling using a scale factor related to the nozzle exit area is typically applied that shifts model scale frequency in proportion to the geometric scale factor. However, model scale DEN designs have two inherent length scales. One is associated with the mini-nozzles, whose size do not change in going from model scale to full scale. The other is associated with the overall nozzle exit area which is much smaller than full size. Consequently, lower frequency energy that is generated by the coalesced jet plume should scale to lower frequency, but higher frequency energy generated by individual mini-jets does not shift frequency. In addition, jet-jet acoustic shielding by the array of mini-nozzles is a significant noise reduction effect that may change with DEN model size. A technique has been developed to scale laboratory model spectral data based on the premise that high and low frequency content must be treated differently during the scaling process. The model-scale distributed exhaust spectra are divided into low and high frequency regions that are then adjusted to full scale separately based on different physics-based scaling laws. The regions are then recombined to create an estimate of the full-scale acoustic spectra. These spectra can then be converted to perceived noise levels (PNL). The paper presents the details of this methodology and provides an example of the estimated noise suppression by a distributed exhaust nozzle compared to a round conic nozzle.

Microphysics in Multi-scale Modeling System with Unified Physics

NASA Technical Reports Server (NTRS)

Tao, Wei-Kuo

2012-01-01

Recently, a multi-scale modeling system with unified physics was developed at NASA Goddard. It consists of (1) a cloud-resolving model (Goddard Cumulus Ensemble model, GCE model), (2) a regional scale model (a NASA unified weather research and forecast, WRF), (3) a coupled CRM and global model (Goddard Multi-scale Modeling Framework, MMF), and (4) a land modeling system. The same microphysical processes, long and short wave radiative transfer and land processes and the explicit cloud-radiation, and cloud-land surface interactive processes are applied in this multi-scale modeling system. This modeling system has been coupled with a multi-satellite simulator to use NASA high-resolution satellite data to identify the strengths and weaknesses of cloud and precipitation processes simulated by the model. In this talk, a review of developments and applications of the multi-scale modeling system will be presented. In particular, the microphysics development and its performance for the multi-scale modeling system will be presented.

Relevance of multiple spatial scales in habitat models: A case study with amphibians and grasshoppers

NASA Astrophysics Data System (ADS)

Altmoos, Michael; Henle, Klaus

2010-11-01

Habitat models for animal species are important tools in conservation planning. We assessed the need to consider several scales in a case study for three amphibian and two grasshopper species in the post-mining landscapes near Leipzig (Germany). The two species groups were selected because habitat analyses for grasshoppers are usually conducted on one scale only whereas amphibians are thought to depend on more than one spatial scale. First, we analysed how the preference to single habitat variables changed across nested scales. Most environmental variables were only significant for a habitat model on one or two scales, with the smallest scale being particularly important. On larger scales, other variables became significant, which cannot be recognized on lower scales. Similar preferences across scales occurred in only 13 out of 79 cases and in 3 out of 79 cases the preference and avoidance for the same variable were even reversed among scales. Second, we developed habitat models by using a logistic regression on every scale and for all combinations of scales and analysed how the quality of habitat models changed with the scales considered. To achieve a sufficient accuracy of the habitat models with a minimum number of variables, at least two scales were required for all species except for Bufo viridis, for which a single scale, the microscale, was sufficient. Only for the European tree frog ( Hyla arborea), at least three scales were required. The results indicate that the quality of habitat models increases with the number of surveyed variables and with the number of scales, but costs increase too. Searching for simplifications in multi-scaled habitat models, we suggest that 2 or 3 scales should be a suitable trade-off, when attempting to define a suitable microscale.

Derivation of a GIS-based watershed-scale conceptual model for the St. Jones River Delaware from habitat-scale conceptual models.

PubMed

Reiter, Michael A; Saintil, Max; Yang, Ziming; Pokrajac, Dragoljub

2009-08-01

Conceptual modeling is a useful tool for identifying pathways between drivers, stressors, Valued Ecosystem Components (VECs), and services that are central to understanding how an ecosystem operates. The St. Jones River watershed, DE is a complex ecosystem, and because management decisions must include ecological, social, political, and economic considerations, a conceptual model is a good tool for accommodating the full range of inputs. In 2002, a Four-Component, Level 1 conceptual model was formed for the key habitats of the St. Jones River watershed, but since the habitat level of resolution is too fine for some important watershed-scale issues we developed a functional watershed-scale model using the existing narrowed habitat-scale models. The narrowed habitat-scale conceptual models and associated matrices developed by Reiter et al. (2006) were combined with data from the 2002 land use/land cover (LULC) GIS-based maps of Kent County in Delaware to assemble a diagrammatic and numerical watershed-scale conceptual model incorporating the calculated weight of each habitat within the watershed. The numerical component of the assembled watershed model was subsequently subjected to the same Monte Carlo narrowing methodology used for the habitat versions to refine the diagrammatic component of the watershed-scale model. The narrowed numerical representation of the model was used to generate forecasts for changes in the parameters "Agriculture" and "Forest", showing that land use changes in these habitats propagated through the results of the model by the weighting factor. Also, the narrowed watershed-scale conceptual model identified some key parameters upon which to focus research attention and management decisions at the watershed scale. The forecast and simulation results seemed to indicate that the watershed-scale conceptual model does lead to different conclusions than the habitat-scale conceptual models for some issues at the larger watershed scale.

Multi-scale modeling in cell biology

PubMed Central

Meier-Schellersheim, Martin; Fraser, Iain D. C.; Klauschen, Frederick

2009-01-01

Biomedical research frequently involves performing experiments and developing hypotheses that link different scales of biological systems such as, for instance, the scales of intracellular molecular interactions to the scale of cellular behavior and beyond to the behavior of cell populations. Computational modeling efforts that aim at exploring such multi-scale systems quantitatively with the help of simulations have to incorporate several different simulation techniques due to the different time and space scales involved. Here, we provide a non-technical overview of how different scales of experimental research can be combined with the appropriate computational modeling techniques. We also show that current modeling software permits building and simulating multi-scale models without having to become involved with the underlying technical details of computational modeling. PMID:20448808

Scale effect challenges in urban hydrology highlighted with a Fully Distributed Model and High-resolution rainfall data

NASA Astrophysics Data System (ADS)

Ichiba, Abdellah; Gires, Auguste; Tchiguirinskaia, Ioulia; Schertzer, Daniel; Bompard, Philippe; Ten Veldhuis, Marie-Claire

2017-04-01

Nowadays, there is a growing interest on small-scale rainfall information, provided by weather radars, to be used in urban water management and decision-making. Therefore, an increasing interest is in parallel devoted to the development of fully distributed and grid-based models following the increase of computation capabilities, the availability of high-resolution GIS information needed for such models implementation. However, the choice of an appropriate implementation scale to integrate the catchment heterogeneity and the whole measured rainfall variability provided by High-resolution radar technologies still issues. This work proposes a two steps investigation of scale effects in urban hydrology and its effects on modeling works. In the first step fractal tools are used to highlight the scale dependency observed within distributed data used to describe the catchment heterogeneity, both the structure of the sewer network and the distribution of impervious areas are analyzed. Then an intensive multi-scale modeling work is carried out to understand scaling effects on hydrological model performance. Investigations were conducted using a fully distributed and physically based model, Multi-Hydro, developed at Ecole des Ponts ParisTech. The model was implemented at 17 spatial resolutions ranging from 100 m to 5 m and modeling investigations were performed using both rain gauge rainfall information as well as high resolution X band radar data in order to assess the sensitivity of the model to small scale rainfall variability. Results coming out from this work demonstrate scale effect challenges in urban hydrology modeling. In fact, fractal concept highlights the scale dependency observed within distributed data used to implement hydrological models. Patterns of geophysical data change when we change the observation pixel size. The multi-scale modeling investigation performed with Multi-Hydro model at 17 spatial resolutions confirms scaling effect on hydrological model performance. Results were analyzed at three ranges of scales identified in the fractal analysis and confirmed in the modeling work. The sensitivity of the model to small-scale rainfall variability was discussed as well.

The sense and non-sense of plot-scale, catchment-scale, continental-scale and global-scale hydrological modelling

NASA Astrophysics Data System (ADS)

Bronstert, Axel; Heistermann, Maik; Francke, Till

2017-04-01

Hydrological models aim at quantifying the hydrological cycle and its constituent processes for particular conditions, sites or periods in time. Such models have been developed for a large range of spatial and temporal scales. One must be aware that the question which is the appropriate scale to be applied depends on the overall question under study. Therefore, it is not advisable to give a general applicable guideline on what is "the best" scale for a model. This statement is even more relevant for coupled hydrological, ecological and atmospheric models. Although a general statement about the most appropriate modelling scale is not recommendable, it is worth to have a look on what are the advantages and the shortcomings of micro-, meso- and macro-scale approaches. Such an appraisal is of increasing importance, since increasingly (very) large / global scale approaches and models are under operation and therefore the question arises how far and for what purposes such methods may yield scientifically sound results. It is important to understand that in most hydrological (and ecological, atmospheric and other) studies process scale, measurement scale, and modelling scale differ from each other. In some cases, the differences between theses scales can be of different orders of magnitude (example: runoff formation, measurement and modelling). These differences are a major source of uncertainty in description and modelling of hydrological, ecological and atmospheric processes. Let us now summarize our viewpoint of the strengths (+) and weaknesses (-) of hydrological models of different scales: Micro scale (e.g. extent of a plot, field or hillslope): (+) enables process research, based on controlled experiments (e.g. infiltration; root water uptake; chemical matter transport); (+) data of state conditions (e.g. soil parameter, vegetation properties) and boundary fluxes (e.g. rainfall or evapotranspiration) are directly measurable and reproducible; (+) equations based on first principals, partly pde-type, are available for several processes (but not for all), because measurement and modelling scale are compatible (-) the spatial model domain are hardly representative for larger spatial entities, including regions for which water resources management decisions are to be taken; straightforward upsizing is also limited by data availability and computational requirements. Meso scale (e.g. extent of a small to large catchment or region): (+) the spatial extent of the model domain has approximately the same extent as the regions for which water resources management decisions are to be taken. I.e., such models enable water resources quantification at the scale of most water management decisions; (+) data of some state conditions (e.g. vegetation cover, topography, river network and cross sections) are available; (+) data of some boundary fluxes (in particular surface runoff / channel flow) are directly measurable with mostly sufficient certainty; (+) equations, partly based on simple water budgeting, partly variants of pde-type equations, are available for most hydrological processes. This enables the construction of meso-scale distributed models reflecting the spatial heterogeneity of regions/landscapes; (-) process scale, measurement scale, and modelling scale differ from each other for a number of processes, e.g., such as runoff generation; (-) the process formulation (usually derived from micro-scale studies) cannot directly be transferred to the modelling domain. Upscaling procedures for this purpose are not readily and generally available. Macro scale (e.g. extent of a continent up to global): (+) the spatial extent of the model may cover the whole Earth. This enables an attractive global display of model results; (+) model results might be technically interchangeable or at least comparable with results from other global models, such as global climate models; (-) process scale, measurement scale, and modelling scale differ heavily from each other for all hydrological and associated processes; (-) the model domain and its results are not representative regions for which water resources management decisions are to be taken. (-) both state condition and boundary flux data are hardly available for the whole model domain. Water management data and discharge data from remote regions are particular incomplete / unavailable for this scale. This undermines the model's verifiability; (-) since process formulation and resulting modelling reliability at this scale is very limited, such models can hardly show any explanatory skills or prognostic power; (-) since both the entire model domain and the spatial sub-units cover large areas, model results represent values averaged over at least the spatial sub-unit's extent. In many cases, the applied time scale implies a long-term averaging in time, too. We emphasize the importance to be aware of the above mentioned strengths and weaknesses of those scale-specific models. (Many of the) results of the current global model studies do not reflect such limitations. In particular, we consider the averaging over large model entities in space and/or time inadequate. Many hydrological processes are of a non-linear nature, including threshold-type behaviour. Such features cannot be reflected by such large scale entities. The model results therefore can be of little or no use for water resources decisions and/or even misleading for public debates or decision making. Some rather newly developed sustainability concepts, e.g. "Planetary Boundaries" in which humanity may "continue to develop and thrive for generations to come" are based on such global-scale approaches and models. However, many of the major problems regarding sustainability on Earth, e.g. water scarcity, do not exhibit on a global but on a regional scale. While on a global scale water might look like being available in sufficient quantity and quality, there are many regions where water problems already have very harmful or even devastating effects. Therefore, it is the challenge to derive models and observation programmes for regional scales. In case a global display is desired future efforts should be directed towards the development of a global picture based on a mosaic of regional sound assessments, rather than "zooming into" the results of large-scale simulations. Still, a key question remains to be discussed, i.e. for which purpose models at this (global) scale can be used.

Concentrating Solar Power Central Receiver Panel Component Fabrication and Testing FINAL REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDowell, Michael W; Miner, Kris

The objective of this project is to complete a design of an advanced concentrated solar panel and demonstrate the manufacturability of key components. Then confirm the operation of the key components under prototypic solar flux conditions. This work is an important step in reducing the levelized cost of energy (LCOE) from a central receiver solar power plant. The key technical risk to building larger power towers is building the larger receiver systems. Therefore, this proposed technology project includes the design of an advanced molten salt prototypic sub-scale receiver panel that can be utilized into a large receiver system. Then completemore » the fabrication and testing of key components of the receive design that will be used to validate the design. This project shall have a significant impact on solar thermal power plant design. Receiver panels of suitable size for utility scale plants are a key element to a solar power tower plant. Many subtle and complex manufacturing processes are involved in producing a reliable, robust receiver panel. Given the substantial size difference between receiver panels manufactured in the past and those needed for large plant designs, the manufacture and demonstration on prototype receiver panel components with representative features of a full-sized panel will be important to improving the build process for commercial success. Given the thermal flux limitations of the test facility, the panel components cannot be rendered full size. Significance changes occurred in the projects technical strategies from project initiation to the accomplishments described herein. The initial strategy was to define cost improvements for the receiver, design and build a scale prototype receiver and test, on sun, with a molten salt heat transport system. DOE had committed to constructing a molten salt heat transport loop to support receiver testing at the top of the NSTTF tower. Because of funding constraints this did not happen. A subsequent plan to test scale prototype receiver, off sun but at temperature, at a molten salt loop at ground level adjacent to the tower also had to be abandoned. Thus, no test facility existed for a molten salt receiver test. As a result, PWR completed the prototype receiver design and then fabricated key components for testing instead of fabricating the complete prototype receiver. A number of innovative design ideas have been developed. Key features of the receiver panel have been identified. This evaluation includes input from Solar 2, personal experience of people working on these programs and meetings with Sandia. Key components of the receiver design and key processes used to fabricate a receiver have been selected for further evaluation. The Test Plan, Concentrated Solar Power Receiver In Cooperation with the Department of Energy and Sandia National Laboratory was written to define the scope of the testing to be completed as well as to provide details related to the hardware, instrumentation, and data acquisition. The document contains a list of test objectives, a test matrix, and an associated test box showing the operating points to be tested. Test Objectives: 1. Demonstrate low-cost manufacturability 2. Demonstrate robustness of two different tube base materials 3. Collect temperature data during on sun operation 4. Demonstrate long term repeated daily operation of heat shields 5. Complete pinhole tube weld repairs 6. Anchor thermal models This report discusses the tests performed, the results, and implications for design improvements and LCOE reduction.« less

Physical consistency of subgrid-scale models for large-eddy simulation of incompressible turbulent flows

NASA Astrophysics Data System (ADS)

Silvis, Maurits H.; Remmerswaal, Ronald A.; Verstappen, Roel

2017-01-01

We study the construction of subgrid-scale models for large-eddy simulation of incompressible turbulent flows. In particular, we aim to consolidate a systematic approach of constructing subgrid-scale models, based on the idea that it is desirable that subgrid-scale models are consistent with the mathematical and physical properties of the Navier-Stokes equations and the turbulent stresses. To that end, we first discuss in detail the symmetries of the Navier-Stokes equations, and the near-wall scaling behavior, realizability and dissipation properties of the turbulent stresses. We furthermore summarize the requirements that subgrid-scale models have to satisfy in order to preserve these important mathematical and physical properties. In this fashion, a framework of model constraints arises that we apply to analyze the behavior of a number of existing subgrid-scale models that are based on the local velocity gradient. We show that these subgrid-scale models do not satisfy all the desired properties, after which we explain that this is partly due to incompatibilities between model constraints and limitations of velocity-gradient-based subgrid-scale models. However, we also reason that the current framework shows that there is room for improvement in the properties and, hence, the behavior of existing subgrid-scale models. We furthermore show how compatible model constraints can be combined to construct new subgrid-scale models that have desirable properties built into them. We provide a few examples of such new models, of which a new model of eddy viscosity type, that is based on the vortex stretching magnitude, is successfully tested in large-eddy simulations of decaying homogeneous isotropic turbulence and turbulent plane-channel flow.

Peridynamic Multiscale Finite Element Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costa, Timothy; Bond, Stephen D.; Littlewood, David John

The problem of computing quantum-accurate design-scale solutions to mechanics problems is rich with applications and serves as the background to modern multiscale science research. The prob- lem can be broken into component problems comprised of communicating across adjacent scales, which when strung together create a pipeline for information to travel from quantum scales to design scales. Traditionally, this involves connections between a) quantum electronic structure calculations and molecular dynamics and between b) molecular dynamics and local partial differ- ential equation models at the design scale. The second step, b), is particularly challenging since the appropriate scales of molecular dynamic andmore » local partial differential equation models do not overlap. The peridynamic model for continuum mechanics provides an advantage in this endeavor, as the basic equations of peridynamics are valid at a wide range of scales limiting from the classical partial differential equation models valid at the design scale to the scale of molecular dynamics. In this work we focus on the development of multiscale finite element methods for the peridynamic model, in an effort to create a mathematically consistent channel for microscale information to travel from the upper limits of the molecular dynamics scale to the design scale. In particular, we first develop a Nonlocal Multiscale Finite Element Method which solves the peridynamic model at multiple scales to include microscale information at the coarse-scale. We then consider a method that solves a fine-scale peridynamic model to build element-support basis functions for a coarse- scale local partial differential equation model, called the Mixed Locality Multiscale Finite Element Method. Given decades of research and development into finite element codes for the local partial differential equation models of continuum mechanics there is a strong desire to couple local and nonlocal models to leverage the speed and state of the art of local models with the flexibility and accuracy of the nonlocal peridynamic model. In the mixed locality method this coupling occurs across scales, so that the nonlocal model can be used to communicate material heterogeneity at scales inappropriate to local partial differential equation models. Additionally, the computational burden of the weak form of the peridynamic model is reduced dramatically by only requiring that the model be solved on local patches of the simulation domain which may be computed in parallel, taking advantage of the heterogeneous nature of next generation computing platforms. Addition- ally, we present a novel Galerkin framework, the 'Ambulant Galerkin Method', which represents a first step towards a unified mathematical analysis of local and nonlocal multiscale finite element methods, and whose future extension will allow the analysis of multiscale finite element methods that mix models across scales under certain assumptions of the consistency of those models.« less

Scale problems in assessment of hydrogeological parameters of groundwater flow models

NASA Astrophysics Data System (ADS)

Nawalany, Marek; Sinicyn, Grzegorz

2015-09-01

An overview is presented of scale problems in groundwater flow, with emphasis on upscaling of hydraulic conductivity, being a brief summary of the conventional upscaling approach with some attention paid to recently emerged approaches. The focus is on essential aspects which may be an advantage in comparison to the occasionally extremely extensive summaries presented in the literature. In the present paper the concept of scale is introduced as an indispensable part of system analysis applied to hydrogeology. The concept is illustrated with a simple hydrogeological system for which definitions of four major ingredients of scale are presented: (i) spatial extent and geometry of hydrogeological system, (ii) spatial continuity and granularity of both natural and man-made objects within the system, (iii) duration of the system and (iv) continuity/granularity of natural and man-related variables of groundwater flow system. Scales used in hydrogeology are categorised into five classes: micro-scale - scale of pores, meso-scale - scale of laboratory sample, macro-scale - scale of typical blocks in numerical models of groundwater flow, local-scale - scale of an aquifer/aquitard and regional-scale - scale of series of aquifers and aquitards. Variables, parameters and groundwater flow equations for the three lowest scales, i.e., pore-scale, sample-scale and (numerical) block-scale, are discussed in detail, with the aim to justify physically deterministic procedures of upscaling from finer to coarser scales (stochastic issues of upscaling are not discussed here). Since the procedure of transition from sample-scale to block-scale is physically well based, it is a good candidate for upscaling block-scale models to local-scale models and likewise for upscaling local-scale models to regional-scale models. Also the latest results in downscaling from block-scale to sample scale are briefly referred to.

Up-scaling of multi-variable flood loss models from objects to land use units at the meso-scale

NASA Astrophysics Data System (ADS)

Kreibich, Heidi; Schröter, Kai; Merz, Bruno

2016-05-01

Flood risk management increasingly relies on risk analyses, including loss modelling. Most of the flood loss models usually applied in standard practice have in common that complex damaging processes are described by simple approaches like stage-damage functions. Novel multi-variable models significantly improve loss estimation on the micro-scale and may also be advantageous for large-scale applications. However, more input parameters also reveal additional uncertainty, even more in upscaling procedures for meso-scale applications, where the parameters need to be estimated on a regional area-wide basis. To gain more knowledge about challenges associated with the up-scaling of multi-variable flood loss models the following approach is applied: Single- and multi-variable micro-scale flood loss models are up-scaled and applied on the meso-scale, namely on basis of ATKIS land-use units. Application and validation is undertaken in 19 municipalities, which were affected during the 2002 flood by the River Mulde in Saxony, Germany by comparison to official loss data provided by the Saxon Relief Bank (SAB).In the meso-scale case study based model validation, most multi-variable models show smaller errors than the uni-variable stage-damage functions. The results show the suitability of the up-scaling approach, and, in accordance with micro-scale validation studies, that multi-variable models are an improvement in flood loss modelling also on the meso-scale. However, uncertainties remain high, stressing the importance of uncertainty quantification. Thus, the development of probabilistic loss models, like BT-FLEMO used in this study, which inherently provide uncertainty information are the way forward.

Large- and small-scale constraints on power spectra in Omega = 1 universes

NASA Technical Reports Server (NTRS)

Gelb, James M.; Gradwohl, Ben-Ami; Frieman, Joshua A.

1993-01-01

The CDM model of structure formation, normalized on large scales, leads to excessive pairwise velocity dispersions on small scales. In an attempt to circumvent this problem, we study three scenarios (all with Omega = 1) with more large-scale and less small-scale power than the standard CDM model: (1) cold dark matter with significantly reduced small-scale power (inspired by models with an admixture of cold and hot dark matter); (2) cold dark matter with a non-scale-invariant power spectrum; and (3) cold dark matter with coupling of dark matter to a long-range vector field. When normalized to COBE on large scales, such models do lead to reduced velocities on small scales and they produce fewer halos compared with CDM. However, models with sufficiently low small-scale velocities apparently fail to produce an adequate number of halos.

Allometric scaling of biceps strength before and after resistance training in men.

PubMed

Zoeller, Robert F; Ryan, Eric D; Gordish-Dressman, Heather; Price, Thomas B; Seip, Richard L; Angelopoulos, Theodore J; Moyna, Niall M; Gordon, Paul M; Thompson, Paul D; Hoffman, Eric P

2007-06-01

The purposes of this study were 1) derive allometric scaling models of isometric biceps muscle strength using pretraining body mass (BM) and muscle cross-sectional area (CSA) as scaling variables in adult males, 2) test model appropriateness using regression diagnostics, and 3) cross-validate the models before and after 12 wk of resistance training. A subset of FAMuSS (Functional SNP Associated with Muscle Size and Strength) study data (N=136) were randomly split into two groups (A and B). Allometric scaling models using pretraining BM and CSA were derived and tested for group A. The scaling exponents determined from these models were then applied to and tested on group B pretraining data. Finally, these scaling exponents were applied to and tested on group A and B posttraining data. BM and CSA models produced scaling exponents of 0.64 and 0.71, respectively. Regression diagnostics determined both models to be appropriate. Cross-validation of the models to group B showed that the BM model, but not the CSA model, was appropriate. Removal of the largest six subjects (CSA>30 cm) from group B resulted in an appropriate fit for the CSA model. Application of the models to group A posttraining data showed that both models were appropriate, but only the body mass model was successful for group B. These data suggest that the application of scaling exponents of 0.64 and 0.71, using BM and CSA, respectively, are appropriate for scaling isometric biceps strength in adult males. However, the scaling exponent using CSA may not be appropriate for individuals with biceps CSA>30 cm. Finally, 12 wk of resistance training does not alter the relationship between BM, CSA, and muscular strength as assessed by allometric scaling.

Novel patch modelling method for efficient simulation and prediction uncertainty analysis of multi-scale groundwater flow and transport processes

NASA Astrophysics Data System (ADS)

Sreekanth, J.; Moore, Catherine

2018-04-01

The application of global sensitivity and uncertainty analysis techniques to groundwater models of deep sedimentary basins are typically challenged by large computational burdens combined with associated numerical stability issues. The highly parameterized approaches required for exploring the predictive uncertainty associated with the heterogeneous hydraulic characteristics of multiple aquifers and aquitards in these sedimentary basins exacerbate these issues. A novel Patch Modelling Methodology is proposed for improving the computational feasibility of stochastic modelling analysis of large-scale and complex groundwater models. The method incorporates a nested groundwater modelling framework that enables efficient simulation of groundwater flow and transport across multiple spatial and temporal scales. The method also allows different processes to be simulated within different model scales. Existing nested model methodologies are extended by employing 'joining predictions' for extrapolating prediction-salient information from one model scale to the next. This establishes a feedback mechanism supporting the transfer of information from child models to parent models as well as parent models to child models in a computationally efficient manner. This feedback mechanism is simple and flexible and ensures that while the salient small scale features influencing larger scale prediction are transferred back to the larger scale, this does not require the live coupling of models. This method allows the modelling of multiple groundwater flow and transport processes using separate groundwater models that are built for the appropriate spatial and temporal scales, within a stochastic framework, while also removing the computational burden associated with live model coupling. The utility of the method is demonstrated by application to an actual large scale aquifer injection scheme in Australia.

Spatial calibration and temporal validation of flow for regional scale hydrologic modeling

USDA-ARS?s Scientific Manuscript database

Physically based regional scale hydrologic modeling is gaining importance for planning and management of water resources. Calibration and validation of such regional scale model is necessary before applying it for scenario assessment. However, in most regional scale hydrologic modeling, flow validat...

Scaling properties of Arctic sea ice deformation in high-resolution viscous-plastic sea ice models and satellite observations

NASA Astrophysics Data System (ADS)

Hutter, Nils; Losch, Martin; Menemenlis, Dimitris

2017-04-01

Sea ice models with the traditional viscous-plastic (VP) rheology and very high grid resolution can resolve leads and deformation rates that are localised along Linear Kinematic Features (LKF). In a 1-km pan-Arctic sea ice-ocean simulation, the small scale sea-ice deformations in the Central Arctic are evaluated with a scaling analysis in relation to satellite observations of the Envisat Geophysical Processor System (EGPS). A new coupled scaling analysis for data on Eulerian grids determines the spatial and the temporal scaling as well as the coupling between temporal and spatial scales. The spatial scaling of the modelled sea ice deformation implies multi-fractality. The spatial scaling is also coupled to temporal scales and varies realistically by region and season. The agreement of the spatial scaling and its coupling to temporal scales with satellite observations and models with the modern elasto-brittle rheology challenges previous results with VP models at coarse resolution where no such scaling was found. The temporal scaling analysis, however, shows that the VP model does not fully resolve the intermittency of sea ice deformation that is observed in satellite data.

The scaling of model test results to predict intake hot gas reingestion for STOVL aircraft with augmented vectored thrust engines

NASA Technical Reports Server (NTRS)

Penrose, C. J.

1987-01-01

The difficulties of modeling the complex recirculating flow fields produced by multiple jet STOVL aircraft close to the ground have led to extensive use of experimental model tests to predict intake Hot Gas Reingestion (HGR). Model test results reliability is dependent on a satisfactory set of scaling rules which must be validated by fully comparable full scale tests. Scaling rules devised in the U.K. in the mid 60's gave good model/full scale agreement for the BAe P1127 aircraft. Until recently no opportunity has occurred to check the applicability of the rules to the high energy exhaust of current ASTOVL aircraft projects. Such an opportunity has arisen following tests on a Tethered Harrier. Comparison of this full scale data and results from tests on a model configuration approximating to the full scale aircraft geometry has shown discrepancies between HGR levels. These discrepancies although probably due to geometry and other model/scale differences indicate some reexamination of the scaling rules is needed. Therefore the scaling rules are reviewed, further scaling studies planned are described and potential areas for further work are suggested.

Multi-Scale Models for the Scale Interaction of Organized Tropical Convection

NASA Astrophysics Data System (ADS)

Yang, Qiu

Assessing the upscale impact of organized tropical convection from small spatial and temporal scales is a research imperative, not only for having a better understanding of the multi-scale structures of dynamical and convective fields in the tropics, but also for eventually helping in the design of new parameterization strategies to improve the next-generation global climate models. Here self-consistent multi-scale models are derived systematically by following the multi-scale asymptotic methods and used to describe the hierarchical structures of tropical atmospheric flows. The advantages of using these multi-scale models lie in isolating the essential components of multi-scale interaction and providing assessment of the upscale impact of the small-scale fluctuations onto the large-scale mean flow through eddy flux divergences of momentum and temperature in a transparent fashion. Specifically, this thesis includes three research projects about multi-scale interaction of organized tropical convection, involving tropical flows at different scaling regimes and utilizing different multi-scale models correspondingly. Inspired by the observed variability of tropical convection on multiple temporal scales, including daily and intraseasonal time scales, the goal of the first project is to assess the intraseasonal impact of the diurnal cycle on the planetary-scale circulation such as the Hadley cell. As an extension of the first project, the goal of the second project is to assess the intraseasonal impact of the diurnal cycle over the Maritime Continent on the Madden-Julian Oscillation. In the third project, the goals are to simulate the baroclinic aspects of the ITCZ breakdown and assess its upscale impact on the planetary-scale circulation over the eastern Pacific. These simple multi-scale models should be useful to understand the scale interaction of organized tropical convection and help improve the parameterization of unresolved processes in global climate models.

Extension of landscape-based population viability models to ecoregional scales for conservation planning

Treesearch

Thomas W. Bonnot; Frank R. III Thompson; Joshua Millspaugh

2011-01-01

Landscape-based population models are potentially valuable tools in facilitating conservation planning and actions at large scales. However, such models have rarely been applied at ecoregional scales. We extended landscape-based population models to ecoregional scales for three species of concern in the Central Hardwoods Bird Conservation Region and compared model...

Biology meets physics: Reductionism and multi-scale modeling of morphogenesis.

PubMed

Green, Sara; Batterman, Robert

2017-02-01

A common reductionist assumption is that macro-scale behaviors can be described "bottom-up" if only sufficient details about lower-scale processes are available. The view that an "ideal" or "fundamental" physics would be sufficient to explain all macro-scale phenomena has been met with criticism from philosophers of biology. Specifically, scholars have pointed to the impossibility of deducing biological explanations from physical ones, and to the irreducible nature of distinctively biological processes such as gene regulation and evolution. This paper takes a step back in asking whether bottom-up modeling is feasible even when modeling simple physical systems across scales. By comparing examples of multi-scale modeling in physics and biology, we argue that the "tyranny of scales" problem presents a challenge to reductive explanations in both physics and biology. The problem refers to the scale-dependency of physical and biological behaviors that forces researchers to combine different models relying on different scale-specific mathematical strategies and boundary conditions. Analyzing the ways in which different models are combined in multi-scale modeling also has implications for the relation between physics and biology. Contrary to the assumption that physical science approaches provide reductive explanations in biology, we exemplify how inputs from physics often reveal the importance of macro-scale models and explanations. We illustrate this through an examination of the role of biomechanical modeling in developmental biology. In such contexts, the relation between models at different scales and from different disciplines is neither reductive nor completely autonomous, but interdependent. Copyright © 2016 Elsevier Ltd. All rights reserved.

Non-linear scaling of a musculoskeletal model of the lower limb using statistical shape models.

PubMed

Nolte, Daniel; Tsang, Chui Kit; Zhang, Kai Yu; Ding, Ziyun; Kedgley, Angela E; Bull, Anthony M J

2016-10-03

Accurate muscle geometry for musculoskeletal models is important to enable accurate subject-specific simulations. Commonly, linear scaling is used to obtain individualised muscle geometry. More advanced methods include non-linear scaling using segmented bone surfaces and manual or semi-automatic digitisation of muscle paths from medical images. In this study, a new scaling method combining non-linear scaling with reconstructions of bone surfaces using statistical shape modelling is presented. Statistical Shape Models (SSMs) of femur and tibia/fibula were used to reconstruct bone surfaces of nine subjects. Reference models were created by morphing manually digitised muscle paths to mean shapes of the SSMs using non-linear transformations and inter-subject variability was calculated. Subject-specific models of muscle attachment and via points were created from three reference models. The accuracy was evaluated by calculating the differences between the scaled and manually digitised models. The points defining the muscle paths showed large inter-subject variability at the thigh and shank - up to 26mm; this was found to limit the accuracy of all studied scaling methods. Errors for the subject-specific muscle point reconstructions of the thigh could be decreased by 9% to 20% by using the non-linear scaling compared to a typical linear scaling method. We conclude that the proposed non-linear scaling method is more accurate than linear scaling methods. Thus, when combined with the ability to reconstruct bone surfaces from incomplete or scattered geometry data using statistical shape models our proposed method is an alternative to linear scaling methods. Copyright © 2016 The Author. Published by Elsevier Ltd.. All rights reserved.

Application of Mortar Coupling in Multiscale Modelling of Coupled Flow, Transport, and Biofilm Growth in Porous Media

NASA Astrophysics Data System (ADS)

Laleian, A.; Valocchi, A. J.; Werth, C. J.

2017-12-01

Multiscale models of reactive transport in porous media are capable of capturing complex pore-scale processes while leveraging the efficiency of continuum-scale models. In particular, porosity changes caused by biofilm development yield complex feedbacks between transport and reaction that are difficult to quantify at the continuum scale. Pore-scale models, needed to accurately resolve these dynamics, are often impractical for applications due to their computational cost. To address this challenge, we are developing a multiscale model of biofilm growth in which non-overlapping regions at pore and continuum spatial scales are coupled with a mortar method providing continuity at interfaces. We explore two decompositions of coupled pore-scale and continuum-scale regions to study biofilm growth in a transverse mixing zone. In the first decomposition, all reaction is confined to a pore-scale region extending the transverse mixing zone length. Only solute transport occurs in the surrounding continuum-scale regions. Relative to a fully pore-scale result, we find the multiscale model with this decomposition has a reduced run time and consistent result in terms of biofilm growth and solute utilization. In the second decomposition, reaction occurs in both an up-gradient pore-scale region and a down-gradient continuum-scale region. To quantify clogging, the continuum-scale model implements empirical relations between porosity and continuum-scale parameters, such as permeability and the transverse dispersion coefficient. Solutes are sufficiently mixed at the end of the pore-scale region, such that the initial reaction rate is accurately computed using averaged concentrations in the continuum-scale region. Relative to a fully pore-scale result, we find accuracy of biomass growth in the multiscale model with this decomposition improves as the interface between pore-scale and continuum-scale regions moves downgradient where transverse mixing is more fully developed. Also, this decomposition poses additional challenges with respect to mortar coupling. We explore these challenges and potential solutions. While recent work has demonstrated growing interest in multiscale models, further development is needed for their application to field-scale subsurface contaminant transport and remediation.

10. MOVABLE BED SEDIMENTATION MODELS. DOGTOOTH BEND MODEL (MODEL SCALE: ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. MOVABLE BED SEDIMENTATION MODELS. DOGTOOTH BEND MODEL (MODEL SCALE: 1' = 400' HORIZONTAL, 1' = 100' VERTICAL), AND GREENVILLE BRIDGE MODEL (MODEL SCALE: 1' = 360' HORIZONTAL, 1' = 100' VERTICAL). - Waterways Experiment Station, Hydraulics Laboratory, Halls Ferry Road, 2 miles south of I-20, Vicksburg, Warren County, MS

Status of DSMT research program

NASA Technical Reports Server (NTRS)

Mcgowan, Paul E.; Javeed, Mehzad; Edighoffer, Harold H.

1991-01-01

The status of the Dynamic Scale Model Technology (DSMT) research program is presented. DSMT is developing scale model technology for large space structures as part of the Control Structure Interaction (CSI) program at NASA Langley Research Center (LaRC). Under DSMT a hybrid-scale structural dynamics model of Space Station Freedom was developed. Space Station Freedom was selected as the focus structure for DSMT since the station represents the first opportunity to obtain flight data on a complex, three-dimensional space structure. Included is an overview of DSMT including the development of the space station scale model and the resulting hardware. Scaling technology was developed for this model to achieve a ground test article which existing test facilities can accommodate while employing realistically scaled hardware. The model was designed and fabricated by the Lockheed Missile and Space Co., and is assembled at LaRc for dynamic testing. Also, results from ground tests and analyses of the various model components are presented along with plans for future subassembly and matted model tests. Finally, utilization of the scale model for enhancing analysis verification of the full-scale space station is also considered.

Investigation of correlation between full-scale and fifth-scale wind tunnel tests of a Bell helicopter Textron Model 222

NASA Technical Reports Server (NTRS)

Squires, P. K.

1982-01-01

Reasons for lack of correlation between data from a fifth-scale wind tunnel test of the Bell Helicopter Textron Model 222 and a full-scale test of the model 222 prototype in the NASA Ames 40-by 80-foot tunnel were investigated. This investigation centered around a carefully designed fifth-scale wind tunnel test of an accurately contoured model of the Model 222 prototype mounted on a replica of the full-scale mounting system. The improvement in correlation for drag characteristics in pitch and yaw with the fifth-scale model mounted on the replica system is shown. Interference between the model and mounting system was identified as a significant effect and was concluded to be a primary cause of the lack of correlation in the earlier tests.

Scaling effects in the static and dynamic response of graphite-epoxy beam-columns. Ph.D. Thesis - Virginia Polytechnic Inst. and State Univ.

NASA Technical Reports Server (NTRS)

Jackson, Karen E.

1990-01-01

Scale model technology represents one method of investigating the behavior of advanced, weight-efficient composite structures under a variety of loading conditions. It is necessary, however, to understand the limitations involved in testing scale model structures before the technique can be fully utilized. These limitations, or scaling effects, are characterized. in the large deflection response and failure of composite beams. Scale model beams were loaded with an eccentric axial compressive load designed to produce large bending deflections and global failure. A dimensional analysis was performed on the composite beam-column loading configuration to determine a model law governing the system response. An experimental program was developed to validate the model law under both static and dynamic loading conditions. Laminate stacking sequences including unidirectional, angle ply, cross ply, and quasi-isotropic were tested to examine a diversity of composite response and failure modes. The model beams were loaded under scaled test conditions until catastrophic failure. A large deflection beam solution was developed to compare with the static experimental results and to analyze beam failure. Also, the finite element code DYCAST (DYnamic Crash Analysis of STructure) was used to model both the static and impulsive beam response. Static test results indicate that the unidirectional and cross ply beam responses scale as predicted by the model law, even under severe deformations. In general, failure modes were consistent between scale models within a laminate family; however, a significant scale effect was observed in strength. The scale effect in strength which was evident in the static tests was also observed in the dynamic tests. Scaling of load and strain time histories between the scale model beams and the prototypes was excellent for the unidirectional beams, but inconsistent results were obtained for the angle ply, cross ply, and quasi-isotropic beams. Results show that valuable information can be obtained from testing on scale model composite structures, especially in the linear elastic response region. However, due to scaling effects in the strength behavior of composite laminates, caution must be used in extrapolating data taken from a scale model test when that test involves failure of the structure.