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

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

Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

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

Vigil, R.A.; Jacobus, M.J.

1994-04-01

Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables,more » the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent.« less

Radiological dose in Muria peninsula from SB-LOCA event

NASA Astrophysics Data System (ADS)

Sunarko; Suud, Zaki

2017-01-01

Dose assessment for accident condition is performed for Muria Peninsula region using source-term from Three-Mile Island unit 2 SB-LOCA accident. Xe-133, Kr-88, 1-131 and Cs-137 isotopes are considered in the calculation. The effluent is assumed to be released from a 50 m stack. Lagrangian particle dispersion method (LPDM) employing non-Gaussian dispersion coefficient in 3-dimensional mass-consistent wind-field is employed to obtain periodic surface-level concentration which is then time-integrated to obtain spatial distribution of ground-level dose. In 1-hour simulation, segmented plumes with 60 seconds duration with a total of 18.000 particles involved. Simulations using 6-hour worst-case meteorological data from Muria peninsula results in a peak external dose of around 1.668 mSv for low scenario and 6.892 mSv for high scenario in dry condition. In wet condition with 5 mm/hour and 10 mm/hour rain for the whole duration of the simulation provides only minor effect to dose. The peak external dose is below the regulatory limit of 50 mSv for effective skin dose from external gamma exposure.

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

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

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

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

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

Szilard, Ronaldo Henriques; Youngblood, Robert; Frepoli, Cesare

2015-04-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review andmore » approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.« less

Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

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

Miao, Yinbin; Ye, Bei; Hofman, Gerard

As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U 3Si 2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U 3Si 2 at LWR conditions. The fission gas behavior of U 3Si 2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranularmore » bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U 3Si 2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U 3Si 2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U 3Si 2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.« 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.

R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

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

Szilard, Ronaldo Henriques; Zhang, Hongbin; Epiney, Aaron Simon

The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licenseemore » costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and

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.

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.

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.

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

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

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.

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.

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

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

Cardoni, Jeffrey

2010-11-01

A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

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.

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

Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

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

Liao, J.; Cao, L.; Ohkawa, K.

2012-07-01

The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model ismore » conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)« 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.

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

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

Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

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

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies willmore » still be present in the successor code RELAP5/MOD3.« 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.

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.

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.

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.

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

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

Baek, J. S.; Cheng, L. Y.; Diamond, D.

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enrichedmore » uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.« 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

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.

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

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

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

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

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

Development of ECT/UT inspection system for bottom mounted instrumentation nozzle of PWR reactor vessels

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

Tanaka, H.; Fukui, S.; Iwahashi, Y.

1994-12-31

The development of inspection technique and tool for Bottom Mounted Instrument (BMI) nozzle of PWR plant was performed for countermeasure of leakage accident at incore instrument nozzle of Hamaoka-1 (BWR). MHI achieved the following development, of which object was PWR Plant R/V: (1) development of ECT/UT Multi-sensored Probe; (2) development of Inspection System (3) development of Data Processing System. The Inspection System had been functionally tested using full scale mock-up. As the result of the functional test, this system was confirmed to be very effective, and assumed to be hopeful for the actual application on site.

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

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

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

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.

Nuclear power plant cable materials :

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

Celina, Mathias C.; Gillen, Kenneth T; Lindgren, Eric Richard

2013-05-01

A selective literature review was conducted to assess whether currently available accelerated aging and original qualification data could be used to establish operational margins for the continued use of cable insulation and jacketing materials in nuclear power plant environments. The materials are subject to chemical and physical degradation under extended radiationthermal- oxidative conditions. Of particular interest were the circumstances under which existing aging data could be used to predict whether aged materials should pass loss of coolant accident (LOCA) performance requirements. Original LOCA qualification testing usually involved accelerated aging simulations of the 40-year expected ambient aging conditions followed by amore » LOCA simulation. The accelerated aging simulations were conducted under rapid accelerated aging conditions that did not account for many of the known limitations in accelerated polymer aging and therefore did not correctly simulate actual aging conditions. These highly accelerated aging conditions resulted in insulation materials with mostly inert aging processes as well as jacket materials where oxidative damage dropped quickly away from the air-exposed outside jacket surface. Therefore, for most LOCA performance predictions, testing appears to have relied upon heterogeneous aging behavior with oxidation often limited to the exterior of the cable cross-section a situation which is not comparable with the nearly homogenous oxidative aging that will occur over decades under low dose rate and low temperature plant conditions. The historical aging conditions are therefore insufficient to determine with reasonable confidence the remaining operational margins for these materials. This does not necessarily imply that the existing 40-year-old materials would fail if LOCA conditions occurred, but rather that unambiguous statements about the current aging state and anticipated LOCA performance cannot be provided based on original

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

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

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

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

Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C

2007-09-01

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally filesmore » and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.« 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

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.

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.

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

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

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

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.

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

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

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

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

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

Church ladies, good girls, and locas: stigma and the intersection of gender, ethnicity, mental illness, and sexuality in relation to HIV risk.

PubMed

Collins, Pamela Y; von Unger, Hella; Armbrister, Adria

2008-08-01

Inner city women with severe mental illness may carry multiple stigmatized statuses. In some contexts these include having a mental illness, being a member of an ethnic minority group, being an immigrant, being poor, and being a woman who does not live up to gendered expectations. These potentially stigmatizing identities influence both the way women's sexuality is viewed and their risk for HIV infection. This qualitative study applies the concept of intersectionality to facilitate understanding of how these multiple identities intersect to influence women's sexuality and HIV risk. We report the firsthand accounts of 24 Latina women living with severe mental illness in New York City. In examining the interlocking domains of these women's sexual lives, we find that the women seek identities that define them in opposition to the stigmatizing label of "loca" (Spanish for crazy) and bestow respect and dignity. These identities have unfolded through the additional themes of "good girls" and "church ladies". Therefore, in spite of their association with the "loca", the women also identify with faith and religion ("church ladies") and uphold more traditional gender norms ("good girls") that are often undermined by the realities of life with a severe mental illness and the stigma attached to it. However, the participants fall short of their gender ideals and engage in sexual relationships that they experience as disempowering and unsatisfying. The effects of their multiple identities as poor Latina women living with severe mental illness in an urban ethnic minority community are not always additive, but the interlocking effects can facilitate increased HIV risks. Interventions should acknowledge women's multiple layers of vulnerability, both individual and structural, and stress women's empowerment in and beyond the sexual realm.

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

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

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.

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

Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

DTIC Science & Technology

1987-06-01

computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

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

Decay Heat Removal from a GFR Core by Natural Convection

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

Williams, Wesley C.; Hejzlar, Pavel; Driscoll, Michael J.

2004-07-01

One of the primary challenges for Gas-cooled Fast Reactors (GFR) is decay heat removal after a loss of coolant accident (LOCA). Due to the fact that thermal gas cooled reactors currently under design rely on passive mechanisms to dissipate decay heat, there is a strong motivation to accomplish GFR core cooling through natural phenomena. This work investigates the potential of post-LOCA decay heat removal from a GFR core to a heat sink using an external convection loop. A model was developed in the form of the LOCA-COLA (Loss of Coolant Accident - Convection Loop Analysis) computer code as a meansmore » for 1D steady state convective heat transfer loop analysis. The results show that decay heat removal by means of gas cooled natural circulation is feasible under elevated post-LOCA containment pressure conditions. (authors)« 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

Irradiation behaviour of the large grained UO{sub 2} fuel pellet in the transient conditions

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

Kosaka, Yuji; Watanabe, Seiichi; Arakawa, Yasushi

2007-07-01

In order to achieve a high duty fuel rod design, it is the key issue to suppress the fission gas release from the view point of the fuel rod inner pressure design. The large grain UO{sub 2} pellet is one of the candidates to meet such a requirement by reducing the fission gas release especially at high power and/or high burnup. We have demonstrated the fuel performance of the large grain pellet in the PWR irradiation conditions, which was fabricated with no additive but with active UO{sub 2} powder through the conventional pelletizing process for the normal grain size pellet.more » According to the mechanism of the fission gas retention, there may be a concern about the larger gas bubble swelling of the large grain pellet at the power transient conditions which may increase the potential of the PCMI failure. In this paper, we focus on the differences of the dimensional change in comparison among the pellets with the different grain sizes at the power transient conditions. The power ramp tests were carried out on the high burnup fuel rods of normal and large grain pellet with no additive, which had been irradiated in the PWR conditions up to around 60 GWd/t at peak position. The detailed PIE results revealed that the volume increment due to the power ramp clearly showed the dependence on the grain size as well as the fission gas release and suggested that the larger grain with no additive may suppress the gas bubble swelling at the power transient conditions. According to the experimental results, it is concluded that the large grain pellet with no additive does not deteriorate the irradiation performance during the power transient conditions from the view point of the gas bubble swelling. (authors)« less

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

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.

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).

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

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

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.

LOFT L2-3 blowdown experiment safety analyses D, E, and G; LOCA analyses H, K, K1

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

Perryman, J.L.; Keeler, C.D.; Saukkoriipi, L.O.

1978-12-01

Three calculations using conservative off-nominal conditions and evaluation model options were made using RELAP4/MOD5 for blowdown-refill and RELAP4/MOD6 for reflood for Loss-of-Fluid Test Experiment L2-3 to support the experiment safety analysis effort. The three analyses are as follows: Analysis D: Loss of commercial power during Experiment L2-3; Analysis E: Hot leg quick-opening blowdown valve (QOBV) does not open during Experiment L2-3; and Analysis G: Cold leg QOBV does not open during Experiment L2-3. In addition, the results of three LOFT loss-of-coolant accident (LOCA) analyses using a power of 56.1 MW and a primary coolant system flow rate of 3.6 millionmore » 1bm/hr are presented: Analysis H: Intact loop 200% hot leg break; emergency core cooling (ECC) system B unavailable; Analysis K: Pressurizer relief valve stuck in open position; ECC system B unavailable; and Analysis K1: Same as analysis K, but using a primary coolant system flow rate of 1.92 million 1bm/hr (L2-4 pre-LOCE flow rate). For analysis D, the maximum cladding temperature reached was 1762/sup 0/F, 22 sec into reflood. In analyses E and G, the blowdowns were slower due to one of the QOBVs not functioning. The maximum cladding temperature reached in analysis E was 1700/sup 0/F, 64.7 sec into reflood; for analysis G, it was 1300/sup 0/F at the start of reflood. For analysis H, the maximum cladding temperature reached was 1825/sup 0/F, 0.01 sec into reflood. Analysis K was a very slow blowdown, and the cladding temperatures followed the saturation temperature of the system. The results of analysis K1 was nearly identical to analysis K; system depressurization was not affected by the primary coolant system flow rate.« less

Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

NASA Astrophysics Data System (ADS)

Artnak, Edward Joseph, III

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

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

Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean

2016-08-01

The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swellingmore » and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.« less

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

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.

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

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

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.

Effects of future climate conditions on terrestrial export from coastal southern California

NASA Astrophysics Data System (ADS)

Feng, D.; Zhao, Y.; Raoufi, R.; Beighley, E.; Melack, J. M.

2015-12-01

The Santa Barbara Coastal - Long Term Ecological Research Project (SBC-LTER) is focused on investigating the relative importance of land and ocean processes in structuring giant kelp forest ecosystems. Understanding how current and future climate conditions influence terrestrial export is a central theme for the project. Here we combine the Hillslope River Routing (HRR) model and daily precipitation and temperature downscaled using statistical downscaling based on localized constructed Analogs (LOCA) to estimate recent streamflow dynamics (2000 to 2014) and future conditions (2015 to 2100). The HRR model covers the SBC-LTER watersheds from just west of the Ventura River to Point Conception; a land area of roughly 800 km2 with 179 watersheds ranging from 0.1 to 123 km2. The downscaled climate conditions have a spatial resolution of 6 km by 6 km. Here, we use the Penman-Monteith method with the Food and Agriculture Organization of the United Nations (FAO) limited climate data approximations and land surface conditions (albedo, leaf area index, land cover) measured from NASA's Moderate Resolution Imaging Spectroradiometer (MODIS) on the Terra and Aqua satellites to estimate potential evapotranspiration (PET). The HRR model is calibrated for the period 2000 to 2014 using USGS and LTER streamflow. An automated calibration technique is used. For future climate scenarios, we use mean 8-day land cover conditions. Future streamflow, ET and soil moisture statistics are presented and based on downscaled P and T from ten climate model projections from the Coupled Model Intercomparison Project Phase 5 (CMIP5).

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

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

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

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

Bias estimates used in lieu of validation of fission products and minor actinides in MCNP K eff calculations for PWR burnup credit casks

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

Mueller, Don E.; Marshall, William J.; Wagner, John C.

The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the biasmore » due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.« less

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.

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

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

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.

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

Geochemical and Hydrologic Controls of Copper-Rich Surface Waters in the Yerba Loca-Mapocho System

NASA Astrophysics Data System (ADS)

Pasten, P.; Montecinos, M.; Coquery, M.; Pizarro, G. E.; Abarca, M. I.; Arce, G. J.

2015-12-01

Andean watersheds in Northern and Central Chile are naturally enriched with metals, many of them associated to sulfide mineralizations related to copper mining districts. The natural and anthropogenic influx of toxic metals into drinking water sources pose a sustainability challenge for cities that need to provide safe water with the smallest footprint. This work presents our study of the transformations of copper in the Yerba Loca-Mapocho system. Our sampling campaign started from the headwaters at La Paloma Glacier and continues to the inlet of the San Enrique drinking water treatment plant, a system feeding municipalities in the Eastern area of Santiago, Chile. Depending on the season, total copper concentrations go as high as 22 mg/L for the upper sections, which become diluted to <5 mg/L downstream. pH ranged from 3 to 5.6 while suspended solids ranged from <10 to 100 mg/L. We used Geochemist Workbench to assess copper speciation and to evaluate the thermodynamic controls for the formation and dissolution of solid phases. A sediment trap was used to concentrate suspended particulate matter, which was analyzed with ICP-MS, TXRF (total reflection X ray fluorescence) and XRD (X-ray diffraction). Major elements detected in the precipitates were Al (200 g/kg), S (60 g/kg), and Cu (6 g/kg). Likely solid phases include hydrous amorphous phases of aluminum hydroxides and sulfates, and copper hydroxides/carbonates. Efforts are undergoing to find the optimal mixing ratios between the acidic stream and more alkaline streams to maximize attenuation of dissolved copper. The results of this research could be used for enhancing in-stream natural attenuation of copper and reducing treatment needs at the drinking water facility. Acknowledgements to Fondecyt 1130936 and Conicyt Fondap 15110020

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

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

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« 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

Pretest analysis of natural circulation on the PWR model PACTEL with horizontal steam generators

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

Kervinen, T.; Riikonen, V.; Ritonummi, T.

A new tests facility - parallel channel tests loop (PACTEL)- has been designed and built to simulate the major components and system behavior of pressurized water reactors (PWRs) during postulated small- and medium-break loss-of-coolant accidents. Pretest calculations have been performed for the first test series, and the results of these calculations are being used for planning experiments, for adjusting the data acquisition system, and for choosing the optimal position and type of instrumentation. PACTEL is a volumetrically scaled (1:305) model of the VVER-440 PWR. In all the calculated cases, the natural circulation was found to be effective in removing themore » heat from the core to the steam generator. The loop mass flow rate peaked at 60% mass inventory. The straightening of the loop seals increased the mass flow rate significantly.« less

MODELLING OF FUEL BEHAVIOUR DURING LOSS-OF-COOLANT ACCIDENTS USING THE BISON CODE

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

Pastore, G.; Novascone, S. R.; Williamson, R. L.

2015-09-01

This work presents recent developments to extend the BISON code to enable fuel performance analysis during LOCAs. This newly developed capability accounts for the main physical phenomena involved, as well as the interactions among them and with the global fuel rod thermo-mechanical analysis. Specifically, new multiphysics models are incorporated in the code to describe (1) transient fission gas behaviour, (2) rapid steam-cladding oxidation, (3) Zircaloy solid-solid phase transition, (4) hydrogen generation and transport through the cladding, and (5) Zircaloy high-temperature non-linear mechanical behaviour and failure. Basic model characteristics are described, and a demonstration BISON analysis of a LWR fuel rodmore » undergoing a LOCA accident is presented. Also, as a first step of validation, the code with the new capability is applied to the simulation of experiments investigating cladding behaviour under LOCA conditions. The comparison of the results with the available experimental data of cladding failure due to burst is presented.« less

Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

NASA Astrophysics Data System (ADS)

Narukawa, Takafumi; Yamaguchi, Akira; Jang, Sunghyon; Amaya, Masaki

2018-02-01

For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions of light-water-reactors, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best among the three models to estimate the fracture probability in terms of the degree of prediction accuracy for both next data to be obtained and the true model. Using the log-probit model, it was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.

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

-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

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

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.

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.

Core cooling under accident conditions at the high flux beam reactor (HFBR)

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

Tichler, P.; Cheng, L.; Fauske, H.

In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. In a LOCA scenario where even limited fuelmore » damage occurs and natural circulation is established, fission product gases could be carried from the damaged fuel by steam into areas where operator access is required to maintain the core in a coolable configuration. This would force evacuation of the building and lead to extensive core damage. As a result the HFBR was shut down by the Department of Energy (DOE) and an extensive review of the HFBR was initiated. In an effort to address this issue BNL developed a model designed to predict the heat removal limit during flow reversal that was found to be in good agreement with the test results. Currently a thermal-hydraulic test program is being developed to provide a more realistic and defensible estimate of the flow reversal heat removal limit so that the reactor power level can be increased.« less

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

Heat, Mass and Aerosol Transfers in Spray Conditions for Containment Application

NASA Astrophysics Data System (ADS)

Porcheron, Emmanuel; Lemaitre, Pascal; Nuboer, Amandine; Vendel, Jacques

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surété Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Post Quench Ductility Evaluation of Zircaloy-4 and Select Iron Alloys under Design Basis and Extended LOCA Conditions

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

Yan, Yong; Keiser, James R; Terrani, Kurt A

2014-01-01

Oxidation experiments were conducted at 1200 C in flowing steam with tubing specimens of Zircaloy-4, 317, 347 stainless steels, and the commercial FeCrAl alloy APMT. The purpose was to determine the oxidation behavior and post quench ductility of these alloys under postulated loss-of-coolant accident conditions. The parabolic rate constant for Zircaloy-4 tubing samples at 1200 were determined to be k = 2.173 107 g2/cm4/s C, in excellent agreement with the Cathcart-Pawel correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. The ductility of post quenched samples was evaluated by ring compression tests atmore » 135 C. For Zircaloy-4, the ductile to brittle transition occurs at an equivalent cladding reacted (ECR) of 19.3%. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR 50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.« 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

Evaluation and comparison of gross primary production estimates for the Northern Great Plains grasslands

USGS Publications Warehouse

Zhang, Li; Wylie, Bruce K.; Loveland, Thomas R.; Fosnight, Eugene A.; Tieszen, Larry L.; Ji, Lei; Gilmanov, Tagir

2007-01-01

Two spatially-explicit estimates of gross primary production (GPP) are available for the Northern Great Plains. An empirical piecewise regression (PWR) GPP model was developed from flux tower measurements to map carbon flux across the region. The Moderate Resolution Imaging Spectrometer (MODIS) GPP model is a process-based model that uses flux tower data to calibrate its parameters. Verification and comparison of the regional PWR GPP and the global MODIS GPP are important for the modeling of grassland carbon flux. This study compared GPP estimates from PWR and MODIS models with five towers in the grasslands. Among them, PWR GPP and MODIS GPP showed a good agreement with tower-based GPP at three towers. The global MODIS GPP, however, did not agree well with tower-based GPP at two other towers, probably because of the insensitivity of MODIS model to regional ecosystem and climate change and extreme soil moisture conditions. Cross-validation indicated that the PWR model is relatively robust for predicting regional grassland GPP. However, the PWR model should include a wide variety of flux tower data as the training data sets to obtain more accurate results.In addition, GPP maps based on the PWR and MODIS models were compared for the entire region. In the northwest and south, PWR GPP was much higher than MODIS GPP. These areas were characterized by the higher water holding capacity with a lower proportion of C4 grasses in the northwest and a higher proportion of C4 grasses in the south. In the central and southeastern regions, PWR GPP was much lower than MODIS GPP under complicated conditions with generally mixed C3/C4 grasses. The analysis indicated that the global MODIS GPP model has some limitations on detecting moisture stress, which may have been caused by the facts that C3 and C4 grasses are not distinguished, water stress is driven by vapor pressure deficit (VPD) from coarse meteorological data, and MODIS land cover data are unable to differentiate the sub

The art in getting flocks and herds to flerds

USDA-ARS?s Scientific Manuscript database

Flerds (small ruminants that consistently stay near cattle under free-ranging conditions) offer four distinct advantages over stocking simply flocks and herds to carry out mixed species stocking. One of the main advantages flerds offer is added protection from canine predation, reduced time in loca...

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.

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

Fuel cladding behavior under rapid loading conditions

NASA Astrophysics Data System (ADS)

Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

2016-02-01

A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

Comparisons of Wilks’ and Monte Carlo Methods in Response to the 10CFR50.46(c) Proposed Rulemaking

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

Zhang, Hongbin; Szilard, Ronaldo; Zou, Ling

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10CFR50.46(c), the US NRC put forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnup-dependent cladding embrittlement criteria; consequently nearly all the fuel rods in a reactor core need to be analyzed under LOCA conditions to demonstrate compliance to the safety limits. New analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of themore » limiting rods as well as to quantify the safety margins under LOCA conditions. With the new analysis method presented in this work, the limiting transient case and the limiting rods can be easily identified to quantify the safety margins in response to the proposed new rulemaking. In this work, the best-estimate plus uncertainty (BEPU) analysis capability for large break LOCA with the new cladding embrittlement criteria using the RELAP5-3D code is established and demonstrated with a reduced set of uncertainty parameters. Both the direct Monte Carlo method and the Wilks’ nonparametric statistical method can be used to perform uncertainty quantification. Wilks’ method has become the de-facto industry standard to perform uncertainty quantification in BEPU LOCA analyses. Despite its widespread adoption by the industry, the use of small sample sizes to infer statement of compliance to the existing 10CFR50.46 rule, has been a major cause of unrealized operational margin in today’s BEPU methods. Moreover the debate on the proper interpretation of the Wilks’ theorem in the context of safety analyses is not fully resolved yet, even more than two decades after its introduction in the frame of safety analyses in the nuclear industry. This represents both a

Development of a Korean reference HLW disposal system under the Korean representative geologic conditions

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

Choi, Heui-Joo; Lee, Jong Youl; Choi, Jongwon

2007-07-01

The development of a Korean Reference disposal System for the spent fuels from PWR and CANDU reactors is outlined in this paper. Around 36,000 tU of spent fuels are being projected based on the lifetimes of 28 nuclear power reactors in Korea. Since the site for the geological disposal has not yet been decided, a hypothetical site with representative Korean geologic conditions is proposed for the conceptual design of the repository. The disposal rates of the spent fuels are determined according to the total operation time of 55 years. The canisters are optimized by considering natural Korean conditions, and themore » buffer is designed with domestic Ca-bentonite. The depth of the repository is determined to be 500 m below the ground's surface. The canister separation distances are determined through a thermal analysis. The main features of the repository are presented from the layout to the closure. A computer program has been developed to calculate and analyze the volume and the area of the disposal system to help in the cost analysis. The final output of the design is presented as a unit disposal cost, US $315 /kgU. (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.

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).

In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

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

Upadhyaya, Belle; Hines, J. Wesley; Damiano, Brian

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. Themore » following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets

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

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

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

Localized Multi-Model Extremes Metrics for the Fourth National Climate Assessment

NASA Astrophysics Data System (ADS)

Thompson, T. R.; Kunkel, K.; Stevens, L. E.; Easterling, D. R.; Biard, J.; Sun, L.

2017-12-01

We have performed localized analysis of scenario-based datasets for the Fourth National Climate Assessment (NCA4). These datasets include CMIP5-based Localized Constructed Analogs (LOCA) downscaled simulations at daily temporal resolution and 1/16th-degree spatial resolution. Over 45 temperature and precipitation extremes metrics have been processed using LOCA data, including threshold, percentile, and degree-days calculations. The localized analysis calculates trends in the temperature and precipitation extremes metrics for relatively small regions such as counties, metropolitan areas, climate zones, administrative areas, or economic zones. For NCA4, we are currently addressing metropolitan areas as defined by U.S. Census Bureau Metropolitan Statistical Areas. Such localized analysis provides essential information for adaptation planning at scales relevant to local planning agencies and businesses. Nearly 30 such regions have been analyzed to date. Each locale is defined by a closed polygon that is used to extract LOCA-based extremes metrics specific to the area. For each metric, single-model data at each LOCA grid location are first averaged over several 30-year historical and future periods. Then, for each metric, the spatial average across the region is calculated using model weights based on both model independence and reproducibility of current climate conditions. The range of single-model results is also captured on the same localized basis, and then combined with the weighted ensemble average for each region and each metric. For example, Boston-area cooling degree days and maximum daily temperature is shown below for RCP8.5 (red) and RCP4.5 (blue) scenarios. We also discuss inter-regional comparison of these metrics, as well as their relevance to risk analysis for adaptation planning.

Hot Cell Installation and Demonstration of the Severe Accident Test Station

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

Linton, Kory D.; Burns, Zachary M.; Terrani, Kurt A.

A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examinemore » postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.« 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

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

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

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

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.

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

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

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

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.

RELAP5/MOD2 analysis of a postulated cold leg SBLOCA'' simultaneous to a total black-out'' event in the Jose Cabrera Nuclear Station

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

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the lessons learned'' in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. The assumptionmore » of a total black-out condition'' coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the time to core overheating startup'' as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

RELAP5/MOD2 analysis of a postulated ``cold leg SBLOCA`` simultaneous to a ``total black-out`` event in the Jose Cabrera Nuclear Station

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

Rebollo, L.

1992-04-01

Several beyond-design bases cold leg small-break LOCA postulated scenarios based on the ``lessons learned`` in the OECD-LOFT LP-SB-3 experiment have been analyzed for the Westinghouse single loop Jose Cabrera Nuclear Power Plant belonging to the Spanish utility UNION ELECTRICA FENOSA, S.A. The analysis has been done by the utility in the Thermal-Hydraulic & Accident Analysis Section of the Engineering Department of the Nuclear Division. The RELAP5/MOD2/36.04 code has been used on a CYBER 180/830 computer and the simulation includes the 6 in. RHRS charging line, the 2 in. pressurizer spray, and the 1.5 in. CVCS make-up line piping breaks. Themore » assumption of a ``total black-out condition`` coincident with the occurrence of the event has been made in order to consider a plant degraded condition with total active failure of the ECCS. As a result of the analysis, estimates of the ``time to core overheating startup`` as well as an evaluation of alternate operator measures to mitigate the consequences of the event have been obtained. Finally a proposal for improving the LOCA emergency operating procedure (E-1) has been suggested.« less

75 FR 53985 - Arizona Public Service Company, et al., Palo Verde Nuclear Generating Station, Unit 3; Temporary...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-02

... are authorized by law, will not present an undue risk to public health or safety, and are consistent... Public Health and Safety The underlying purpose of 10 CFR 50.46 is to establish acceptance criteria for... (LOCA) and non-LOCA criteria, mechanical design, thermal hydraulics, seismic, core physics, and...

Transport and mixing of a volume of fluid in a complex geometry

NASA Astrophysics Data System (ADS)

Gavelli, Filippo

This work presents the results of the experimental investigation of an entire sequence of events, leading to an unwanted injection of boron-depleted water into the core of a PWR. The study is subdivided into three tasks: the generation of a dilute volume in the primary system, its transport to the core, and the mixing encountered along the path. Experiments conducted at the University of Maryland (UM) facility show that, during a Small-Break LOCA transient, volumes of dilute coolant are segregated in the system, by means of phase-separating energy transport from the core to the steam generators (Boiler Condenser Mode). Two motion-initiating mechanisms are considered: the resumption of natural circulation during the recovery of the primary liquid inventory, and the reactor coolant pump startup under BCM conditions. During the inventory recovery, various phenomena are observed, that contribute to the mixing of the dilute volumes prior to the resumption of flow. The pump activation, instead, occurs in a stagnant system, therefore, no mixing of the unborated liquid has occurred. Since an unmixed slug has the potential for a larger reactivity excursion than a partially mixed one, the pump-initiated flow resumption represents the worst-case scenario. The impulse - response method is applied, for the first time, to the problem of mixing in the downcomer. This allows to express the mixing in terms of two parameters, the dispersion number and the residence time, characteristics of the flow distribution in the complex annular geometry. Other important results are obtained from the analysis of the experimental data with this procedure. It is shown that the turbulence generated by the pump impeller has a significant impact on the overall mixing. Also, the geometric discontinuities in the downcomer (in particular, the gap enlargement below the cold leg elevation) are shown to be the cause of vortex structures that highly enhance the mixing process.

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.

75 FR 8139 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-23

... the large break loss-of-coolant accident (LOCA) analysis methodology with a reference to WCAP-16009-P... required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards... Section 5.6.5 to incorporate a new large break LOCA analysis methodology. Specifically, the proposed...

Study of clad ballooning and rupture behaviour of Indian PHWR fuel pins under transient heating condition in steam environment

NASA Astrophysics Data System (ADS)

Sawarn, Tapan K.; Banerjee, Suparna; Sheelvantra, Smita S.; Singh, J. L.; Bhasin, Vivek

2017-11-01

This paper presents the results of the investigation on the deformation and rupture characteristics of Indian pressurized heavy water reactor (IPHWR) fuel pins under simulated loss of coolant accident (LOCA) condition in steam environment. Transient heating experiments were carried out on single fuel pin internally pressurized with argon gas in the range 3-70 bar. Effect of internal pressure on burst temperature, influence of burst temperature on the circumferential strain and rupture opening area were also studied. Two circumferential strain maxima at the burst temperatures of 740 & ∼979 °C and a minimum at the burst temperature of ∼868 °C were observed. It was found that oxidation had considerable effect on the burst behavior. Test data were used to derive a direct empirical correlation for burst stress exclusively as a function of temperature. The ballooning and rupture behaviours in steam and argon environments have been compared. Experimental data were examined against various correlations using Erbacher equation and author's previous correlation in argon. A second burst correlation has also been developed combining the equation in argon from the previous work of the authors and an exponential factor with oxygen content as a parameter assuming the burst stress to be a function of both temperature and oxygen concentration. The burst temperatures predicted by this empirical correlation are in good agreement with the test data.

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.

Recent plant studies using Victoria 2.0

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

BIXLER,NATHAN E.; GASSER,RONALD D.

2000-03-08

VICTORIA 2.0 is a mechanistic computer code designed to analyze fission product behavior within the reactor coolant system (RCS) during a severe nuclear reactor accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS and secondary circuits. These predictions account for the chemical and aerosol processes that affect radionuclide behavior. VICTORIA 2.0 was released in early 1999; a new version VICTORIA 2.1, is now under development. The largest improvements in VICTORIA 2.1 are connected with the thermochemical database, which is being revised andmore » expanded following the recommendations of a peer review. Three risk-significant severe accident sequences have recently been investigated using the VICTORIA 2.0 code. The focus here is on how various chemistry options affect the predictions. Additionally, the VICTORIA predictions are compared with ones made using the MELCOR code. The three sequences are a station blackout in a GE BWR and steam generator tube rupture (SGTR) and pump-seal LOCA sequences in a 3-loop Westinghouse PWR. These sequences cover a range of system pressures, from fully depressurized to full system pressure. The chief results of this study are the fission product fractions that are retained in the core, RCS, secondary, and containment and the fractions that are released into the environment.« less

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

NASA Astrophysics Data System (ADS)

Katsuyama, Jinya; Uno, Shumpei; Watanabe, Tadashi; Li, Yinsheng

2018-03-01

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

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.

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.

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

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...

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

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.

Development of Facility Type Information Packages for Design of Air Force Facilities.

DTIC Science & Technology

1983-03-01

solution. For example, the optimum size and loca- 19 tion of windows for the incorporation of a passive solar *l . heating system varies with location, time...conditioning load estimate M. Energy impact statement N. Majcom review comments 0. Solar energy systems 61 4 Information which could help in the development...and Passive solar systems. All facilities should have Scme aspects of passive solar incor- por3ted into the iesign. Active sclar systems should ze con

Core cooling under accident conditions at the high-flux beam reactor

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

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« 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.

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

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

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

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

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

None

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

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...

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

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

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

Hursin, M.; Perret, G.

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handlingmore » and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)« less

Pressure suppression system

DOEpatents

Gluntz, Douglas M.

1994-01-01

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Severe Accident Test Station Design Document

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

Snead, Mary A.; Yan, Yong; Howell, Michael

The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure tomore » provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.« less

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...

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...

Pressure suppression system

DOEpatents

Gluntz, D.M.

1994-10-04

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein. 3 figs.

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.

Pressure suppression containment system

DOEpatents

Gluntz, Douglas M.; Townsend, Harold E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto.

Pressure suppression containment system

DOEpatents

Gluntz, D.M.; Townsend, H.E.

1994-03-15

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of-coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto. 6 figures.

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

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

Ishii, Mamoru

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

Experimental study of Siphon breaker about size effect in real scale reactor design

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

Kang, S. H.; Ahn, H. S.; Kim, J. M.

2012-07-01

Rupture accident within the pipe of a nuclear reactor is one of the main causes of a loss of coolant accident (LOCA). Siphon-breaking is a passive method that can prevent a LOCA. In this study, either a line or a hole is used as a siphon-breaker, and the effect of various parameters, such as the siphon-breaker size, pipe rupture point, pipe rupture size, and the presence of an orifice, are investigated using an experimental facility similar in size to a full-scale reactor. (authors)

Low Platelet to White Blood Cell Ratio Indicates Poor Prognosis for Acute-On-Chronic Liver Failure.

PubMed

Jie, Yusheng; Gong, Jiao; Xiao, Cuicui; Zhu, Shuguang; Zhou, Wenying; Luo, Juan; Chong, Yutian; Hu, Bo

2018-01-01

Background. Platelet to white blood cell ratio (PWR) was an independent prognostic predictor for outcomes in some diseases. However, the prognostic role of PWR is still unclear in patients with hepatitis B related acute-on-chronic liver failure (ACLF). In this study, we evaluated the clinical performances of PWR in predicting prognosis in HBV-related ACLF. Methods. A total of 530 subjects were recruited, including 97 healthy controls and 433 with HBV-related ACLF. Liver function, prothrombin time activity (PTA), international normalized ratio (INR), HBV DNA measurement, and routine hematological testing were performed at admission. Results . At baseline, PWR in patients with HBV-related ACLF (14.03 ± 7.17) was significantly decreased compared to those in healthy controls (39.16 ± 9.80). Reduced PWR values were clinically associated with the severity of liver disease and the increased mortality rate. Furthermore, PWR may be an inexpensive, easily accessible, and significant independent prognostic index for mortality on multivariate analysis (HR = 0.660, 95% CI: 0.438-0.996, p = 0.048) as well as model for end-stage liver disease (MELD) score. Conclusions . The PWR values were markedly decreased in ACLF patients compared with healthy controls and associated with severe liver disease. Moreover, PWR was an independent prognostic indicator for the mortality rate in patients with ACLF. This investigation highlights that PWR comprised a useful biomarker for prediction of liver severity.

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

Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

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

Szilard, Ronaldo H.; Youngblood, Robert W.; Zhang, Hongbin

2015-09-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions formore » NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.« 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

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.

Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance

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

Williford, R.E.

1986-09-01

Current emergency core cooling system acceptance criteria for light water reactors specify that, under loss-of-coolant accident (LOCA) conditions, the Baker-Just (BJ) correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (ECR). An appropriately defined minimum margin of safety was estimated for each of these criteria. The currently required BJ oxidation correlation provides margins only over the 1100 to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperaturesmore » above 1204/sup 0/C for up to 210 and 160 s, respectively, at the 95% confidence level. The ECR thermal shock and handling margins at the 50 and 95% confidence levels, respectively, range between 2 and 7% ECR for the BJ correlation, but vanish at temperatures above 1100 to 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. However, use of the Cathcart Pawel correlation for ''design basis'' LOCA calculations can be justified at the 85 to 88% confidence level if cooling rate effects can be neglected.« less

A review for identification of initiating events in event tree development process on nuclear power plants

NASA Astrophysics Data System (ADS)

Riyadi, Eko H.

2014-09-01

Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.

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

Barry, Kenneth

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

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...

Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA

NASA Astrophysics Data System (ADS)

Yadav, Ashwini Kumar; kumar, Ravi; Gupta, Akhilesh; Chatterjee, Barun; Mukhopadhyay, Deb; Lele, H. G.

2014-06-01

In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.

TRAC analyses for CCTF and SCTF tests and UPTF design/operation. [Cylindrical Core Test Facility; Slab Core Test Facility; Upper Plenum Test Facility

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

Spore, J.W.; Cappiello, M.W.; Dotson, P.J.

The analytical support in 1985 for Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF), and Upper Plenum Test Facility (UPTF) tests involves the posttest analysis of 16 tests that have already been run in the CCTF and the SCTF and the pretest analysis of 3 tests to be performed in the UPTF. Posttest analysis is used to provide insight into the detailed thermal-hydraulic phenomena occurring during the refill and reflood tests performed in CCTF and SCTF. Pretest analysis is used to ensure that the test facility is operated in a manner consistent with the expected behavior of anmore » operating full-scale plant during an accident. To obtain expected behavior of a plant during an accident, two plant loss-of-coolant-accident (LOCA) calculations were performed: a 200% cold-leg-break LOCA calculation for a 2772 MW(t) Babcock and Wilcox plant and a 200% cold-leg-break LOCA calculation for a 3315 MW(t) Westinghouse plant. Detailed results are presented for several CCTF UPI tests and the Westinghouse plant analysis.« less

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

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

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

Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.

1998-04-01

For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

A review for identification of initiating events in event tree development process on nuclear power plants

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

Riyadi, Eko H., E-mail: e.riyadi@bapeten.go.id

2014-09-30

Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logicmore » model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.« less

Multi level optimization of burnable poison utilization for advanced PWR fuel management

NASA Astrophysics Data System (ADS)

Yilmaz, Serkan

The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

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

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

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).

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

Emergy assessment of three home courtyard agriculture production systems in Tibet Autonomous Region, China*

PubMed Central

Guan, Fa-chun; Sha, Zhi-peng; Zhang, Yu-yang; Wang, Jun-feng; Wang, Chao

2016-01-01

Home courtyard agriculture is an important model of agricultural production on the Tibetan plateau. Because of the sensitive and fragile plateau environment, it needs to have optimal performance characteristics, including high sustainability, low environmental pressure, and high economic benefit. Emergy analysis is a promising tool for evaluation of the environmental-economic performance of these production systems. In this study, emergy analysis was used to evaluate three courtyard agricultural production models: Raising Geese in Corn Fields (RGICF), Conventional Corn Planting (CCP), and Pea-Wheat Rotation (PWR). The results showed that the RGICF model produced greater economic benefits, and had higher sustainability, lower environmental pressure, and higher product safety than the CCP and PWR models. The emergy yield ratio (EYR) and emergy self-support ratio (ESR) of RGICF were 0.66 and 0.11, respectively, lower than those of the CCP production model, and 0.99 and 0.08, respectively, lower than those of the PWR production model. The impact of RGICF (1.45) on the environment was lower than that of CCP (2.26) and PWR (2.46). The emergy sustainable indices (ESIs) of RGICF were 1.07 and 1.02 times higher than those of CCP and PWR, respectively. With regard to the emergy index of product safety (EIPS), RGICF had a higher safety index than those of CCP and PWR. Overall, our results suggest that the RGICF model is advantageous and provides higher environmental benefits than the CCP and PWR systems. PMID:27487808

Emergy assessment of three home courtyard agriculture production systems in Tibet Autonomous Region, China.

PubMed

Guan, Fa-Chun; Sha, Zhi-Peng; Zhang, Yu-Yang; Wang, Jun-Feng; Wang, Chao

2016-08-01

Home courtyard agriculture is an important model of agricultural production on the Tibetan plateau. Because of the sensitive and fragile plateau environment, it needs to have optimal performance characteristics, including high sustainability, low environmental pressure, and high economic benefit. Emergy analysis is a promising tool for evaluation of the environmental-economic performance of these production systems. In this study, emergy analysis was used to evaluate three courtyard agricultural production models: Raising Geese in Corn Fields (RGICF), Conventional Corn Planting (CCP), and Pea-Wheat Rotation (PWR). The results showed that the RGICF model produced greater economic benefits, and had higher sustainability, lower environmental pressure, and higher product safety than the CCP and PWR models. The emergy yield ratio (EYR) and emergy self-support ratio (ESR) of RGICF were 0.66 and 0.11, respectively, lower than those of the CCP production model, and 0.99 and 0.08, respectively, lower than those of the PWR production model. The impact of RGICF (1.45) on the environment was lower than that of CCP (2.26) and PWR (2.46). The emergy sustainable indices (ESIs) of RGICF were 1.07 and 1.02 times higher than those of CCP and PWR, respectively. With regard to the emergy index of product safety (EIPS), RGICF had a higher safety index than those of CCP and PWR. Overall, our results suggest that the RGICF model is advantageous and provides higher environmental benefits than the CCP and PWR systems.

Development and Implementation of Mechanistic Terry Turbine Models in RELAP-7 to Simulate RCIC Normal Operation Conditions

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

Zhao, Haihua; Zou, Ling; Zhang, Hongbin

As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC (Reactor Core Isolation Cooling) systems in Fukushima accidents and extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine, based on Sandia’s original work [1], have been developed and implemented in the RELAP-7 code to simulate the RCIC system. In 2016, our effort has been focused on normal working conditions of the RCIC system. More complex off-design conditions will be pursued in later years when more data are available. In the Sandia model, the turbine stator inletmore » velocity is provided according to a reduced-order model which was obtained from a large number of CFD (computational fluid dynamics) 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 stator inlet. The models include both an adiabatic expansion process inside the nozzle and a free expansion process outside of the nozzle to 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 information for the Terry turbine rotor model. The analytical models for the nozzle 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 newly developed nozzle models and modified turbine rotor model according to the Sandia’s original work have been implemented into RELAP-7, along with the original Sandia Terry turbine model. A new pump model has also been developed and implemented to couple with the Terry turbine model. An

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

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

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.

Mechanism-based modeling of solute strengthening: application to thermal creep in Zr alloy

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

Tome, Carlos; Wen, Wei; Capolungo, Laurent

2017-08-01

This report focuses on the development of a physics-based thermal creep model aiming to predict the behavior of Zr alloy under reactor accident condition. The current models used for this kind of simulations are mostly empirical in nature, based generally on fits to the experimental steady-state creep rates under different temperature and stress conditions, which has the following limitations. First, reactor accident conditions, such as RIA and LOCA, usually take place in short times and involve only the primary, not the steady-state creep behavior stage. Moreover, the empirical models cannot cover the conditions from normal operation to accident environments. Formore » example, Kombaiah and Murty [1,2] recently reported a transition between the low (n~4) and high (n~9) power law creep regimes in Zr alloys depending on the applied stress. Capturing such a behavior requires an accurate description of the mechanisms involved in the process. Therefore, a mechanism-based model that accounts for the evolution with time of microstructure is more appropriate and reliable for this kind of simulation.« less

A Study of the Jettisoning of JP-4 Fuel in the Atmosphere

DTIC Science & Technology

1975-11-01

t e r m i n e d by m e a s u r i n g the s tagna t ion p r e s s u r e with a pitot probe , shown in Fig. 21, with the s p h e r i c a l t...ip. The rod loca ted to the r igh t of the pitot probe was loca ted in the s a m e photographic plane as the suspended J P - 4 drop- le t...of Atomiza t ion in C a r b u r e t o r s . " NACA TM 518, 1929. 13. Lapple , C. E . , Henry , J . P . , J r . , and Blake, D. E. " A t o

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

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

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.

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.

Energy conditions and junction conditions

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

Marolf, Donald; Yaida, Sho; Mathematics Department, UCSB, Santa Barbara, California 93106

2005-08-15

We consider the familiar junction conditions described by Israel for thin timelike walls in Einstein-Hilbert gravity. One such condition requires the induced metric to be continuous across the wall. Now, there are many spacetimes with sources confined to a thin wall for which this condition is violated and the Israel formalism does not apply. However, we explore the conjecture that the induced metric is in fact continuous for any thin wall which models spacetimes containing only positive energy matter. Thus, the usual junction conditions would hold for all positive energy spacetimes. This conjecture is proven in various special cases, includingmore » the case of static spacetimes with spherical or planar symmetry as well as settings without symmetry which may be sufficiently well approximated by smooth spacetimes with well-behaved null geodesic congruences.« 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.

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

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

Franken, Daniel; Gould, Daniel; Jain, Prashant K.

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE PAGES

Franken, Daniel; Gould, Daniel; Jain, Prashant K.; ...

2017-09-21

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

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

A guide to aviation education resources

DOT National Transportation Integrated Search

1993-01-01

The National Coalition for Aviation Education represents industry and labor, united to promote : aviation education activities and resources; increase public understanding of the importance of aviation; and support educational initiatives at the loca...

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

Incubation of conditioned fear in the conditioned suppression model in rats: role of food-restriction conditions, length of conditioned stimulus, and generality to conditioned freezing

PubMed Central

Pickens, Charles L.; Navarre, Brittany M.; Nair, Sunila G.

2010-01-01

We recently adapted the conditioned suppression of operant responding method to study fear incubation. We found that food-restricted rats show low fear 2 days after extended (10 d; 100 30-sec tone-shock pairings) fear training and high fear after 1–2 months. Here, we studied a potential mechanism of fear incubation: extended food-restriction stress. We also studied whether fear incubation is observed after fear training with a prolonged-duration (6-min) tone conditioned stimulus (CS), and whether conditioned freezing incubates after extended training in rats with or without a concurrent operant task. Conditioned fear was assessed 2 days and 1 month after training. In the conditioned suppression method, fear incubation was reliably observed in rats under moderate food-restriction conditions (18–20 g food/day) that allowed for weight gain, and after extended (10 d), but not limited (1 d), fear training with the 6-min CS. Incubation of conditioned freezing was observed after extended fear training in rats lever-pressing for food and, to a lesser degree, in rats not performing an operant task. Results indicate that prolonged hunger-related stress does not account for fear incubation in the conditioned suppression method, and that fear incubation occurs to a longer-duration (6-min) fear CS. Extended training also leads to robust fear incubation of conditioned freezing in rats performing an operant task and weaker fear incubation in rats not performing an operant task. PMID:20600654

Incubation of conditioned fear in the conditioned suppression model in rats: role of food-restriction conditions, length of conditioned stimulus, and generality to conditioned freezing.

PubMed

Pickens, C L; Navarre, B M; Nair, S G

2010-09-15

We recently adapted the conditioned suppression of operant responding method to study fear incubation. We found that food-restricted rats show low fear 2 days after extended (10 d; 100 30-s tone-shock pairings) fear training and high fear after 1-2 months. Here, we studied a potential mechanism of fear incubation: extended food-restriction stress. We also studied whether fear incubation is observed after fear training with a prolonged-duration (6-min) tone conditioned stimulus (CS), and whether conditioned freezing incubates after extended training in rats with or without a concurrent operant task. Conditioned fear was assessed 2 days and 1 month after training. In the conditioned suppression method, fear incubation was reliably observed in rats under moderate food-restriction conditions (18-20 g food/day) that allowed for weight gain, and after extended (10 d), but not limited (1 d), fear training with the 6-min CS. Incubation of conditioned freezing was observed after extended fear training in rats lever-pressing for food and, to a lesser degree, in rats not performing an operant task. Results indicate that prolonged hunger-related stress does not account for fear incubation in the conditioned suppression method, and that fear incubation occurs to a longer-duration (6-min) fear CS. Extended training also leads to robust fear incubation of conditioned freezing in rats performing an operant task and weaker fear incubation in rats not performing an operant task. (c) 2010 IBRO. Published by Elsevier Ltd. All rights reserved.

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

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

Joy L. Rempe; Darrell L. Knudson

2013-03-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensorsmore » that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

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

Joy L. Rempe; Darrell L. Knudson

2014-05-01

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensorsmore » that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data

Innovative tools and techniques in identifying highway safety improvement projects : project summary.

DOT National Transportation Integrated Search

2017-01-01

Researchers completed the following activities: - Reviewed the literature, state HSIP processes and practices, and HSIP tools used by various agencies. - Evaluated the applicability of safety assessment methods and tools used by other states and loca...

Joy Development Properties, LLC, Pleasant Valley, Iowa and Summit Concrete, Inc., LeClaire, Iowa - Clean Water Act Public Notice

EPA Pesticide Factsheets

The EPA is providing notice of a proposed Administrative Penalty Assessment against Joy Development Properties, LLC and Summit Concrete, Inc., for alleged violations at the companies’ residential construction site known as the Schutter Farms Addition loca

Partnership strategies for safety roadside rest areas.

DOT National Transportation Integrated Search

2009-01-01

This project studied the many factors influencing the potential for public private partnerships for Safety : Roadside Rest Areas. It found that Federal and California State laws and regulations represent important : barriers to certain types and loca...

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.

Vehicle kinematics in turns and the role of cornering lamps in driver vision.

DOT National Transportation Integrated Search

2010-11-01

"SAE Recommended Practice J852 and ECE Regulations 119 and 48 for cornering lamps : were compared. Photometric points described in each specification were then compared : to naturalistic low-speed turn trajectories produced by 87 drivers. Future loca...

A framework for collaboration in public transit systems

DOT National Transportation Integrated Search

1997-05-01

The 494 transportation corridor stretches eight miles and connects residential suburbs with major commercial areas, including the Mall of America and the Minneapolis-St. Paul International Airport. The corridor includes l-494 as well as parallel loca...

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.

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

Children with chronic conditions: perspectives on condition management.

PubMed

Beacham, Barbara L; Deatrick, Janet A

2015-01-01

This qualitative study described children's (8-13 years old) perspectives of their chronic health conditions (e.g., asthma, diabetes, cystic fibrosis): how they perceived their condition, its management, and its implications for their future. The study used the family management style framework (FMSF) to examine child perspectives on the joint venture of condition management between the child and family. Children within this age group viewed condition management in ways similar to their parents and have developed their own routines around condition management. Future studies of this phenomenon comparing child and parent perspectives would further our understanding of the influence of family management. Copyright © 2015 Elsevier Inc. All rights reserved.

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

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

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

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.

Effects of short-term tocopherol (T) feeding on structure-localized protein tyrosine nitration (pTN) patterns of mitochondrial ATPase following endotoxin (LPS) challenge in beef calves.

USDA-ARS?s Scientific Manuscript database

Mitochondrial ATPase/Complex-V (MCV) is an electron transport chain (ETC) component needed for ATP synthesis. The ETC, exquisitely sensitive to proinflammatory mediators (PIM), generates oxynitrogen reactants leading to pTN formation as mitochondrial membrane leakage occurs. Immunohistochemical loca...

AROMATASE ACTIVITY IN THE OVARY OF MOSQUITOFISH GAMBUSIA HOLBROOKI, COLLECTED FROM THE FENHOLLOWAY AND ECONFINA RIVERS, FLORIDA (

EPA Science Inventory

Scientists are increasingly aware of the adverse effects of environmental contaminants, including their ability to alter the normal development and reproduction of wildlife species by modifying the endocrine system. Female mosquitofish living downstream of a paper mill plant loca...

VHF-FM Emergency Position Indicating Radio Beacon

DOT National Transportation Integrated Search

1978-10-01

This report describes the development and testing of an Emergency Position Indicating Radio Beacon (EPIRB) which operates on Channels 15 and 16 of the Maritime Mobile VHF Band. It provides functions necessary to ensure that distress alerting and loca...

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...

APT Blanket Thermal Analyses of Top Horizontal Row 1 Modules

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

Shadday, M.A.

1999-09-20

The Accelerator Production of Tritium (APT) cavity flood system (CFS) is designed to be the primary safeguard for the integrity of the blanket modules during loss of coolant accidents (LOCAs). For certain large break LOCAs the CFS also provides backup for the residual heat removal systems (RHRs) in cooling the target assemblies. In the unlikely event that the internal flow passages in a blanket module or target assembly dryout, decay heat in the metal structures will be dissipated to the CFS through the module or assembly walls (i.e., rung outer walls). The target assemblies consist of tungsten targets encased inmore » steel conduits, and they can safely sustain high metal temperatures. Under internally dry conditions, the cavity flood fluid will cool the target assemblies with vigorous nucleate boiling on the external surfaces. However, the metal structures in the blanket modules consist of lead cladded in aluminum, and they have a long-term exposure temperature limit currently set to 150 degrees C. Simultaneous LOCAs in both the target and blanket heat removal systems (HRS) could result in dryout of the target ladders, as well as the horizontal blanket modules above the target. The cavity flood coolant would boil on the outside surfaces of the target ladder rungs, and the resultant steam could reduce the effectiveness of convection heat transfer from the blanket modules to the cavity flood coolant. A two-part analysis was conducted to ascertain if the cavity flood system can adequately cool the blanket modules above the targets, even when boiling is occurring on the outer surfaces of the target ladder rungs. The first part of the analysis was to model transient thermal conduction in the front top horizontal row 1 module (i.e. top horizontal modules nearest the incoming beam), while varying parametrically the convection heat transfer coefficient (htc) for the external surfaces exposed to the cavity flood flow. This part of the analysis demonstrated that the

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.

Traffic control at stop sign approaches.

DOT National Transportation Integrated Search

2003-04-01

The objectives of this report were to: a) determine the number of crashes in Kentucky involving a driver disregarding a stop sign and the locations where these occur, b) determine the characteristics of these crashes, c) investigate loca tions with a...

Serotonin and conditioning: focus on Pavlovian psychostimulant drug conditioning.

PubMed

Carey, Robert J; Damianopoulos, Ernest N

2015-04-01

Serotonin containing neurons are located in nuclei deep in the brainstem and send axons throughout the central nervous system from the spinal cord to the cerebral cortex. The vast scope of these connections and interactions enable serotonin and serotonin analogs to have profound effects upon sensory/motor processes. In that conditioning represents a neuroplastic process that leads to new sensory/motor connections, it is apparent that the serotonin system has the potential for a critical role in conditioning. In this article we review the basics of conditioning as well as the serotonergic system and point up the number of non-associative ways in which manipulations of serotonin neurotransmission have an impact upon conditioning. We focus upon psychostimulant drug conditioning and review the contribution of drug stimuli in the use of serotonin drugs to investigate drug conditioning and the important impact drug stimuli can have on conditioning by introducing new sensory stimuli that can create or mask a CS. We also review the ways in which experimental manipulations of serotonin can disrupt conditioned behavioral effects but not the associative processes in conditioning. In addition, we propose the use of the recently developed memory re-consolidation model of conditioning as an approach to assess the possible role of serotonin in associative processes without the complexities of performance effects related to serotonin treatment induced alterations in sensory/motor systems. Copyright © 2014 Elsevier B.V. All rights reserved.

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.

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« 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.

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

A combination of sexual and ecological divergence contributes to the spread of a chromosomal rearrangement during initial stages of speciation

USDA-ARS?s Scientific Manuscript database

Chromosomal rearrangements between sympatric species often contain multiple loci contributing to assortative mating, local adaptation, and hybrid sterility. When and how these associations arise during the process of speciation remains a subject of debate. Here, we address the relative roles of loca...

Passive containment cooling system

DOEpatents

Billig, P.F.; Cooke, F.E.; Fitch, J.R.

1994-01-25

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA. 1 figure.

Passive containment cooling system

DOEpatents

Billig, Paul F.; Cooke, Franklin E.; Fitch, James R.

1994-01-01

A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA.

The negated conditional: a litmus test for the suppositional conditional?

PubMed

Handley, Simon J; Evans, Jonathan St B T; Thompson, Valerie A

2006-05-01

Under the suppositional account of conditionals, when people think about a conditional assertion, "if p then q," they engage in a mental simulation in which they imagine p holds and evaluate the probability that q holds under this supposition. One implication of this account is that belief in a conditional equates to conditional probability [P(q/p)]. In this paper, the authors examine a further implication of this analysis with respect to the wide-scope negation of conditional assertions, "it is not the case that if p then q." Under the suppositional account, nothing categorically follows from the negation of a conditional, other than a second conditional, "if p then not-q." In contrast, according to the mental model theory, a negated conditional is consistent only with the determinate state of affairs, p and not-q. In 4 experiments, the authors compare the contrasting predictions that arise from each of these accounts. The findings are consistent with the suppositional theory but are incongruent with the mental model theory of conditionals.

Polarity in Conditionals and Conditional-Like Constructions

ERIC Educational Resources Information Center

Hsieh, I-Ta Chris

2012-01-01

This dissertation concerns the distribution of negative polarity items (henceforth, NPIs) in conditionals and conditional-like constructions. NPIs include words such as any and ever and idioms such as "give a damn" and "lift a finger"; these expressions have only a limited distribution. In this dissertation, the distribution of…

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

Chinese Conditionals and the Theory of Conditionals.

ERIC Educational Resources Information Center

Chierchia, Gennaro

2000-01-01

Summarizes the main features of discourse representation theory, situation-based approaches, and dynamic semantics, and discusses the role the novelty condition plays in each of them, providing the main theoretical coordinates against which to try an assessment of the role of the Chinese conditional and a proposal put forth by Chang and Huang…

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.

Evaluation of a barley core collection for spot form net blotch reaction reveals distinct genotype specific pathogen virulence and host susceptibility

USDA-ARS?s Scientific Manuscript database

Spot form net blotch (SFNB) caused by Pyrenophora teres Drechs. f. maculata Smedeg., (anamorph Drechslera teres [Sacc.] Shoem.) is a major foliar disease of barley (Hordeum vulgare L.) worldwide. SFNB epidemics have recently been observed in major barley producing countries, suggesting that the loca...

Effect of sampling location on L* values and pH measurements and their relationship in broiler breast fillets

USDA-ARS?s Scientific Manuscript database

Lightness (CIELAB L*) and pH values are the most widely measured quality indicators for broiler breast fillets (pectoralis major). Measurement of L* values with a spectrophotometer can be done through Specular Component Included (SCI) or Specular Component Excluded (SCE) modes. The intra-fillet loca...

The Negated Conditional: A Litmus Test for the Suppositional Conditional?

ERIC Educational Resources Information Center

Handley, Simon J.; Evans, Jonathan St. B. T.; Thompson, Valerie A.

2006-01-01

Under the suppositional account of conditionals, when people think about a conditional assertion, "if p then q," they engage in a mental simulation in which they imagine p holds and evaluate the probability that q holds under this supposition. One implication of this account is that belief in a conditional equates to conditional probability…

Classic conditioning in aged rabbits: delay, trace, and long-delay conditioning.

PubMed

Solomon, P R; Groccia-Ellison, M E

1996-06-01

Young (0.5 years) and aged (2+, 3+, and 4+ years) rabbits underwent acquisition of the classically conditioned nictitating membrane response in a delay (500-ms conditioned stimulus [CS], 400-ms interstimulus interval [ISI]), long-delay (1,000-ms CS, 900-ms ISI), or trace (500-ms CS, 400-ms stimulus-free period) paradigm. Collapsing across age groups, there is a general tendency for animals to acquire trace conditioning more slowly than delay conditioning. Collapsing across conditioning paradigms, there is a general tendency for aged animals to acquire more slowly than younger animals. Of greater significance, however, are the age differences in the different conditioning paradigms. In the delay and long-delay paradigms, significant conditioning deficits first appeared in the 4(+)-year-old group. In the trace conditioning paradigm, significant conditioning deficits became apparent in the 2(+)-year-old animals.

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

Overview of experimental support for fission-product transport analyses at Oak Ridge National Laboratory

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

Wichner, R.P.

The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions with the data from related experiments. These tests of PWR fuel were conducted and fuel specimen and test operating data are presented. The nature and rate of fission product vapor interaction with aerosols were studied. Aerosol deposition rates and transport in the reactor vessel during LWR core-melt accidents were studied. The Nuclear Safety Pilot Plant is dedicated to developing an expanded data basemore » on the behavior of aerosols generated during a severe accident.« less

Characterization and corrosion behavior of F6NM stainless steel treated in high temperature water

NASA Astrophysics Data System (ADS)

Li, Zheng-yang; Cai, Zhen-bing; Yang, Wen-jin; Shen, Xiao-yao; Xue, Guo-hong; Zhu, Min-hao

2018-03-01

F6NM martensitic stainless steel was exposed to 350 °C water condition for 500, 1500, and 2500 h to simulate pressurized water reactor (PWR) condition. The characterization and corrosion behavior of the oxide film were investigated. Results indicate that the exposed steel surface formed a double-layer oxide film. The outer oxide film is Fe-rich and contains two type oxide particles. However, the inner oxide film is Cr-rich, and two oxide films, whose thicknesses increase with increasing exposure time. The oxide film reduces the corrosion behavior because the outer oxide film has many crack and pores. Finally, the mechanism and factors affecting the formation of the oxide film were investigated.

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

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

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

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

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)

Conditional uncertainty principle

NASA Astrophysics Data System (ADS)

Gour, Gilad; Grudka, Andrzej; Horodecki, Michał; Kłobus, Waldemar; Łodyga, Justyna; Narasimhachar, Varun

2018-04-01

We develop a general operational framework that formalizes the concept of conditional uncertainty in a measure-independent fashion. Our formalism is built upon a mathematical relation which we call conditional majorization. We define conditional majorization and, for the case of classical memory, we provide its thorough characterization in terms of monotones, i.e., functions that preserve the partial order under conditional majorization. We demonstrate the application of this framework by deriving two types of memory-assisted uncertainty relations, (1) a monotone-based conditional uncertainty relation and (2) a universal measure-independent conditional uncertainty relation, both of which set a lower bound on the minimal uncertainty that Bob has about Alice's pair of incompatible measurements, conditioned on arbitrary measurement that Bob makes on his own system. We next compare the obtained relations with their existing entropic counterparts and find that they are at least independent.

False Memories for Shape Activate the Lateral Occipital Complex

ERIC Educational Resources Information Center

Karanian, Jessica M.; Slotnick, Scott D.

2017-01-01

Previous functional magnetic resonance imaging evidence has shown that false memories arise from higher-level conscious processing regions rather than lower-level sensory processing regions. In the present study, we assessed whether the lateral occipital complex (LOC)--a lower-level conscious shape processing region--was associated with false…

Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

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

Gonzalez Gonzalez, R.; Petruzzi, A.; D'Auria, F.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D{sup C} system code. Within the framework of themore » third Agreement 'NA-SA - Univ. of Pisa' a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions. (authors)« less

Experimental study of phase separation in dividing two phase flow

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

Qian Yong; Yang Zhilin; Xu Jijun

1996-12-31

Experimental study of phase separation of air-water two phase bubbly, slug flow in the horizontal T-junction is carried out. The influences of the inlet mass quality X1, mass extraction rate G3/G1, and fraction of extracted liquid QL3/QL1 on phase separation characteristics are analyzed. For the first time, the authors have found and defined pulsating run effect by the visual experiments, which show that under certain conditions, the down stream flow of the T-junction has strangely affected the phase redistribution of the junction, and firstly point out that the downstream geometric condition is very important to the study of phase separationmore » phenomenon of two-phase flow in a T-junction. This kind of phenomenon has many applications in the field of energy, power, petroleum and chemical industries, such as the loss of coolant accident (LOCA) caused by a small break in a horizontal coolant pipe in nuclear reactor, and the flip-flop effect in the natural gas transportation pipeline system, etc.« less

Benign Breast Conditions

MedlinePlus

... common benign breast condition in men is called gynecomastia. This condition causes enlarged breast tissue. Female breasts ... with these less common, benign breast conditions. Male gynecomastia: A man’s breast will feel swollen and tender ...

Ecological Condition

EPA Pesticide Factsheets

The ROE is divided into 5 themes: Air, Water, Land, Human Exposure and Health and Ecological Condition. From these themes, the report indicators address fundamental questions that the ROE attempts to answer. For ecological condition there are 5 questions.

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.

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

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

Imaging Near-Earth Electron Densities Using Thomson Scattering

DTIC Science & Technology

2009-01-15

geocentric solar magnetospheric (GSM) coordinates1. TECs were initially computed from a viewing loca- tion at the Sun-Earth L1 Lagrange point2 for both...further find that an elliptical Earth orbit (apogee ~30 RE) is a suitable lower- cost option for a demonstration mission. 5. SIMULATED OBSERVATIONS We

Nuclear safety. Technical progress journal, October 1996--December 1996

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

None

The five papers in this issue address various issues associated with the behavior of high burnup fuels, especially under reactivity initiated accident (RIA) conditions. The mechanisms and parameters that have an effect on the fuel behavior are detailed, based on tests and analyses. The ultimate goal of the research reported is the development of new regulatory criteria for high burnup fuel under design basis accident conditions. Specific topics of the papers, which are abstracted individually in the database, are: (1) regulatory assessment of test data for RIAs, (2) high burnup fuel transient behavior under RIA conditions, (3) NSRR/RIA experiments withmore » high burnup PWR fuels, (4) the Russian RIA research program, and (5) RIA simulation experiments on the intermediate and high burnup test rods. The papers are contributed from the United States, France, Japan, and Russia.« less

Relationship between Air Pollution and Weather Conditions under Complicated Geographical conditions

NASA Astrophysics Data System (ADS)

Cheng, Q.; Jiang, P.; Li, M.

2017-12-01

Air pollution is one of the most serious issues all over the world, especially in megacities with constrained geographical conditions for air pollution diffusion. However, the dynamic mechanism of air pollution diffusion under complicated geographical conditions is still be confused. Researches to explore relationship between air pollution and weather conditions from the perspective of local atmospheric circulations can contribute more to solve such problem. We selected three megacities (Beijing, Shanghai and Guangzhou) under different geographical condition (mountain-plain transition region, coastal alluvial plain and coastal hilly terrain) to explore the relationship between air pollution and weather conditions. RDA (Redundancy analysis) model was used to analyze how the local atmospheric circulation acts on the air pollutant diffusion. The results show that there was a positive correlation between the concentration of air pollutants and air pressure, while temperature, precipitation and wind speed have negative correlations with the concentration of air pollutants. Furthermore, geographical conditions, such as topographic relief, have significant effects on the direction, path and intensity of local atmospheric circulation. As a consequence, air pollutants diffusion modes in different cities under various geographical conditions are diverse from each other.

Fixed or adapted conditioning intensity for repeated conditioned pain modulation.

PubMed

Hoegh, M; Petersen, K K; Graven-Nielsen, T

2017-12-29

Aims Conditioned pain modulation (CPM) is used to assess descending pain modulation through a test stimulation (TS) and a conditioning stimulation (CS). Due to potential carry-over effects, sequential CPM paradigms might alter the intensity of the CS, which potentially can alter the CPM-effect. This study aimed to investigate the difference between a fixed and adaptive CS intensity on CPM-effect. Methods On the dominant leg of 20 healthy subjects the cuff pressure detection threshold (PDT) was recorded as TS and the pain tolerance threshold (PTT) was assessed on the non-dominant leg for estimating the CS. The difference in PDT before and during CS defined the CPM-effect. The CPM-effect was assessed four times using a CS with intensities of 70% of baseline PTT (fixed) or 70% of PTT measured throughout the session (adaptive). Pain intensity of the conditioning stimulus was assessed on a numeric rating scale (NRS). Data were analyzed with repeated-measures ANOVA. Results No difference was found comparing the four PDTs assessed before CSs for the fixed and the adaptive paradigms. The CS pressure intensity for the adaptive paradigm was increasing during the four repeated assessments (P < 0.01). The pain intensity was similar during the fixed (NRS: 5.8±0.5) and the adjusted paradigm (NRS: 6.0±0.4). The CPM-effect was higher using the fixed condition compared with the adaptive condition (P < 0.05). Conclusions The current study found that sequential CPM paradigms using a fixed conditioning stimulus produced an increased CPM-effect compared with adaptive and increasing conditioning intensities.

Crown Condition

Treesearch

KaDonna C. Randolph

2009-01-01

Photosynthetic capacity is dependent upon the size and condition of the tree crown. Trees with full, vigorous crowns are generally associated with more vigorous growth rates (Zarnoch and others 2004). Therefore, the Forest Service Forest Inventory and Analysis (FIA) Program measures a suite of crown condition indicators to evaluate forest health. Among the crown...

Patient to Health Team Communications Preferences and Perceptions of Secure Messaging

DTIC Science & Technology

2017-04-25

Ellicott C itv MD, 25- 27 April 2017 in accordance with MDWI 4 1- 108, has been approved and assigned loca l fi le # 17202. 2. Pe11 inent biographic...scholarl y activities o f our professional staff and students, which is an essential component of Wi lford Hall Ambulatory Surgical Center (WHASC

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

Dissociating contingency awareness and conditioned attitudes: evidence of contingency-unaware evaluative conditioning.

PubMed

Hütter, Mandy; Sweldens, Steven; Stahl, Christoph; Unkelbach, Christian; Klauer, Karl Christoph

2012-08-01

Whether human evaluative conditioning can occur without contingency awareness has been the subject of an intense and ongoing debate for decades, troubled by a wide array of methodological difficulties. Following recent methodological innovations, the available evidence currently points to the conclusion that evaluative conditioning effects do not occur without contingency awareness. In a simulation, we demonstrate, however, that these innovations are strongly biased toward the conclusion that evaluative conditioning requires contingency awareness, confounding the measurement of contingency memory with conditioned attitudes. We adopt a process-dissociation procedure to separate the memory and attitude components. In 4 studies, the attitude parameter is validated using existing attitudes and applied to probe for contingency-unaware evaluative conditioning. A fifth experiment incorporates a time-delay manipulation confirming the dissociability of the attitude and memory components. The results indicate that evaluative conditioning can produce attitudes without conscious awareness of the contingencies. Implications for theories of evaluative conditioning and associative learning are discussed. (PsycINFO Database Record (c) 2012 APA, all rights reserved).

REACH. Air Conditioning Units.

ERIC Educational Resources Information Center

Garrison, Joe; And Others

As a part of the REACH (Refrigeration, Electro-Mechanical, Air-Conditioning, Heating) electromechanical cluster, this student manual contains individualized instructional units in the area of air conditioning. The instructional units focus on air conditioning fundamentals, window air conditioning, system and installation, troubleshooting and…

Pretest and posttest calculations of Semiscale Test S-07-10D with the TRAC computer program. [PWR

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

Duerre, K.H.; Cort, G.E.; Knight, T.D.

The Transient Reactor Analysis Code (TRAC) developed at the Los Alamos National Laboratory was used to predict the behavior of the small-break experiment designated Semiscale S-07-10D. This test simulates a 10 per cent communicative cold-leg break with delayed Emergency Core Coolant injection and blowdown of the broken-loop steam generator secondary. Both pretest calculations that incorporated measured initial conditions and posttest calculations that incorporated measured initial conditions and measured transient boundary conditions were completed. The posttest calculated parameters were generally between those obtained from pretest calculations and those from the test data. The results are strongly dependent on depressurization rate and,more » hence, on break flow.« less

Effect of excess dietary salt on calcium metabolism and bone mineral in a spaceflight rat model

NASA Technical Reports Server (NTRS)

Navidi, Meena; Wolinsky, Ira; Fung, Paul; Arnaud, Sara B.

1995-01-01

High levels of salt promote urinary calcium (UCa) loss and have the potential to cause bone mineral deficits if intestinal Ca absorption does not compensate for these losses. To determine the effect of excess dietary salt on the osteopenia that follows skeletal unloading, we used a spaceflight model that unloads the hindlimbs of 200-g rats by tail suspension (S). Rats were studied for 2 wk on diets containing high salt (4 and 8%) and normal calcium (0.45%) and for 4 wk on diets containing 8% salt (HiNa) and 0.2% Ca (LoCa). Final body weights were 9-11% lower in S than in control rats (C) in both experiments, reflecting lower growth rates in S than in C during pair feeding. UCa represented 12% of dietary Ca on HiNA diets and was twofold higher in S than in C transiently during unloading. Net intestinal Ca absorption was consistently 11-18% lower in S than in C. Serum 1,25-dihydroxyvitamin D was unaffected by either LoCa or HiNa diets in S but was increased by LoCa and HiNa diets in C. Despite depressed intestinal Ca absoption in S and a sluggish response of the Ca endocrine system to HiNa diets, UCa loss did not appear to affect the osteopenia induced by unloading. Although any deficit in bone mineral content from HiNa diets may have been too small to detect or the duration of the study too short to manifest, there were clear differences in Ca metabolism from control levels in the response of the spaceflight model to HiNa diets, indicated by depression of intestinal Ca absorption and its regulatory hormone.

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

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.

[Conditioning mechanisms and psychoneuroimmunology].

PubMed

Stockhorst, Ursula; Klosterhalfen, Sibylle

2005-01-01

This chapter deals with the role of conditioning principles in psychoneuroimmunology (PNI). We will first describe the paradigms of classical and instrumental conditioning and classify immune parameters that are subject to conditioning (chapter 1). So far, PNI research mainly uses classical (or Pavlovian) conditioning. We will summarize some of the paradigmatic studies, mainly animal studies (chapter 2) and also describe studies that support the clinical relevance of classical conditioning, i. e., in the pharmacological treatment of autoimmune diseases, transplantation and tumor chemotherapy (chapter 3). A study of our group on anticipatory immunomodulation in pediatric cancer patients is reported. Mechanisms mediating conditioned immunomodulation are summarized (chapter 4). We also describe studies that analyze the impact of instrumental conditioning contingencies on immune functioning (chapter 5). Finally, research perspectives are summarized (chapter 6).

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.

Generalized Rainich conditions, generalized stress-energy conditions, and the Hawking-Ellis classification

NASA Astrophysics Data System (ADS)

Martín–Moruno, Prado; Visser, Matt

2017-11-01

The (generalized) Rainich conditions are algebraic conditions which are polynomial in the (mixed-component) stress-energy tensor. As such they are logically distinct from the usual classical energy conditions (NEC, WEC, SEC, DEC), and logically distinct from the usual Hawking-Ellis (Segré-Plebański) classification of stress-energy tensors (type I, type II, type III, type IV). There will of course be significant inter-connections between these classification schemes, which we explore in the current article. Overall, we shall argue that it is best to view the (generalized) Rainich conditions as a refinement of the classical energy conditions and the usual Hawking-Ellis classification.

How People Interpret Conditionals: Shifts toward the Conditional Event

ERIC Educational Resources Information Center

Fugard, Andrew J. B.; Pfeifer, Niki; Mayerhofer, Bastian; Kleiter, Gernot D.

2011-01-01

We investigated how people interpret conditionals and how stable their interpretation is over a long series of trials. Participants were shown the colored patterns on each side of a 6-sided die and were asked how sure they were that a conditional holds of the side landing upward when the die is randomly thrown. Participants were presented with 71…

Oxidation of 304 stainless steel in high-temperature steam

NASA Astrophysics Data System (ADS)

Ishida, Toshihisa; Harayama, Yasuo; Yaguchi, Sinnosuke

1986-08-01

An experiment on oxidation of 304 stainless steel was performed in steam between 900°C and 1350°C, using the spare cladding of the reactor of the nuclear-powered ship Mutsu. The temperature range was appropriate for a postulated loss of coolant accident (LOCA) analysis of a LWR. The oxidation kinetics were found to obey the parabolic law during the first period of 8 min. After the first period, the parabolic reaction rate constant decreased in the case of heating temperatures between 1100°C and 1250°C. At 1250°C, especially, a marked decrease was observed in the oxide scale-forming kinetics when the surface treated initially by mechanical polishing and given a residual stress. This enhanced oxidation resistance was attributed to the presence of a chromium-enriched layer which was detected by use of an X-ray microanalyzer. The oxidation kinetics equation obtained for the first 8 min is applicable to the model calculation of a hypothetical LOCA in a LWR, employing 304 stainless steel cladding.

CRACK GROWTH RESPONSE OF ALLOY 690 IN SIMULATED PWR PRIMARY WATER

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

Toloczko, Mychailo B.; Bruemmer, Stephen M.

2009-12-01

The stress corrosion crack growth response of three extruded alloy 690 CRDM tube heats was investigated in several thermomechanical conditions. Extremely low propagation rates (< 1 x 10{sup -9} mm/s) were observed under constant stress intensity factor (K) loading at 325-350 C in the as-received, thermally treated (TT) materials despite using a variety of transitioning techniques. Post-test observation of the crack-growth surfaces revealed only isolated intergranular (IG) cracking. One-dimensional cold rolling to 17% reduction and testing in the S-L orientation did not promote enhanced stress corrosion rates. However, somewhat higher propagation rates were observed in a 30% cold-rolled alloy 690TTmore » specimen tested in the T-L orientation. Cracking of the cold-rolled material was promoted on grain boundaries oriented parallel to the rolling plane with the % IG increasing with the amount of cold rolling.« less

Effects of the air–steam mixture on the permeability of damaged concrete

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

Medjigbodo, Sonagnon; Darquennes, Aveline; Aubernon, Corentin

Massive concrete structures such as the containments of nuclear power plant must maintain their tightness at any circumstances to prevent the escape of radioactive fission products into the environment. In the event of an accident like a Loss of Coolant Accident (LOCA), the concrete wall is submitted to both hydric and mechanical loadings. A new experimental device reproducing these extreme conditions (water vapor transfer, 140 °C and 5 bars) is developed in the GeM Laboratory to determine the effect of the saturation degree, the mechanical loading and the flowing fluid type on the concrete transfer properties. The experimental tests showmore » that the previous parameters significantly affect the concrete permeability and the gas leakage rate. Their evolution as a function of the mechanical loading is characterized by two phases that are directly related to concrete microstructure and crack development.« less

A strategy for oxygen conditioning at high altitude: comparison with air conditioning.

PubMed

West, John B

2015-09-15

Large numbers of people live or work at high altitude, and many visit to trek or ski. The inevitable hypoxia impairs physical working capacity, and at higher altitudes there is also cognitive impairment. Twenty years ago oxygen enrichment of room air was introduced to reduce the hypoxia, and this is now used in dormitories, hotels, mines, and telescopes. However, recent advances in technology now allow large amounts of oxygen to be obtained from air or cryogenic oxygen sources. As a result it is now feasible to oxygenate large buildings and even institutions such as hospitals. An analogy can be drawn between air conditioning that has improved the living and working conditions of millions of people who live in hot climates and oxygen conditioning that can do the same at high altitude. Oxygen conditioning is similar to air conditioning except that instead of cooling the air, the oxygen concentration is raised, thus reducing the equivalent altitude. Oxygen conditioning on a large scale could transform living and working conditions at high altitude, where it could be valuable in homes, hospitals, schools, dormitories, company headquarters, banks, and legislative settings. Copyright © 2015 the American Physiological Society.

Conditional Poisson models: a flexible alternative to conditional logistic case cross-over analysis.

PubMed

Armstrong, Ben G; Gasparrini, Antonio; Tobias, Aurelio

2014-11-24

The time stratified case cross-over approach is a popular alternative to conventional time series regression for analysing associations between time series of environmental exposures (air pollution, weather) and counts of health outcomes. These are almost always analyzed using conditional logistic regression on data expanded to case-control (case crossover) format, but this has some limitations. In particular adjusting for overdispersion and auto-correlation in the counts is not possible. It has been established that a Poisson model for counts with stratum indicators gives identical estimates to those from conditional logistic regression and does not have these limitations, but it is little used, probably because of the overheads in estimating many stratum parameters. The conditional Poisson model avoids estimating stratum parameters by conditioning on the total event count in each stratum, thus simplifying the computing and increasing the number of strata for which fitting is feasible compared with the standard unconditional Poisson model. Unlike the conditional logistic model, the conditional Poisson model does not require expanding the data, and can adjust for overdispersion and auto-correlation. It is available in Stata, R, and other packages. By applying to some real data and using simulations, we demonstrate that conditional Poisson models were simpler to code and shorter to run than are conditional logistic analyses and can be fitted to larger data sets than possible with standard Poisson models. Allowing for overdispersion or autocorrelation was possible with the conditional Poisson model but when not required this model gave identical estimates to those from conditional logistic regression. Conditional Poisson regression models provide an alternative to case crossover analysis of stratified time series data with some advantages. The conditional Poisson model can also be used in other contexts in which primary control for confounding is by fine

Manpower Requirements Report for FY 1982

DTIC Science & Technology

1981-02-01

Specifically included are program elements for industrial preparedness, second destination transportation, property disposal, production engineering ...artillery, and combat - engineers . Army policy accepts the fact that women will serve in loca- .. tions throughout the battlefield, will be expected to... industrial engineering work measurement techniques and computerized models such as the Logistics Composite Model (LCOM). MEP policy emanates from the

Multitarget Tracking Studies.

DTIC Science & Technology

1981-07-01

ie "n.: r c:.t. ’ur wn Le r":2- ence Indistes :nac pole a n :arc loca- 40 ":Lors !AVe a sinificant effect n convergence raes.- .ahdn te poles and...LATTICE ALGORITHMS ( LADO ) 1. The Normalized AR Lattice (ARN) This algorithm implements the normalized algorithm described in [23]. The AR coefficients are

77 FR 53923 - Biweekly Notice;

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-04

... to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well... (psig) to 49.7 psig for the design basis loss-of- coolant accident (LOCA). In support of the revised P a... analysis. The P a remains below the containment design pressure of 50 psig because of the change in the...

Modern Data Analysis techniques in Noise and Vibration Problems

DTIC Science & Technology

1981-11-01

Hilbert l’une de l’autre. Cette propriete se retrouve dans l’etude de la causalite : ce qui de- finit un critere pratique caracterisant un signal donc, par...entre Ie champ direct et Ie champ reflechi se caracterisent loca- lement par l’existence de frequences pour lesquelles l’interference est totale

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

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.

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).

Condition interference in rats performing a choice task with switched variable- and fixed-reward conditions.

PubMed

Funamizu, Akihiro; Ito, Makoto; Doya, Kenji; Kanzaki, Ryohei; Takahashi, Hirokazu

2015-01-01

Because humans and animals encounter various situations, the ability to adaptively decide upon responses to any situation is essential. To date, however, decision processes and the underlying neural substrates have been investigated under specific conditions; thus, little is known about how various conditions influence one another in these processes. In this study, we designed a binary choice task with variable- and fixed-reward conditions and investigated neural activities of the prelimbic cortex and dorsomedial striatum in rats. Variable- and fixed-reward conditions induced flexible and inflexible behaviors, respectively; one of the two conditions was randomly assigned in each trial for testing the possibility of condition interference. Rats were successfully conditioned such that they could find the better reward holes of variable-reward-condition and fixed-reward-condition trials. A learning interference model, which updated expected rewards (i.e., values) used in variable-reward-condition trials on the basis of combined experiences of both conditions, better fit choice behaviors than conventional models which updated values in each condition independently. Thus, although rats distinguished the trial condition, they updated values in a condition-interference manner. Our electrophysiological study suggests that this interfering value-updating is mediated by the prelimbic cortex and dorsomedial striatum. First, some prelimbic cortical and striatal neurons represented the action-reward associations irrespective of trial conditions. Second, the striatal neurons kept tracking the values of variable-reward condition even in fixed-reward-condition trials, such that values were possibly interferingly updated even in the fixed-reward condition.

Road condition reporting.

DOT National Transportation Integrated Search

2015-01-01

This report documents the findings of the road condition reporting project where the feasibility of live reporting of the road : conditions with an Android camera and computer vision algorithms was tested. An app was developed that can collect videos...

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

Enhancing GDOT's Computerized Pavement Condition Evaluation System for Pavement Condition Survey

DOT National Transportation Integrated Search

2017-09-01

The Computerized Pavement Condition Evaluation System (COPACES) is a software tool that has been used by the Georgia Department of Transportation (GDOT) for its statewide pavement condition survey since the late 1990s. The previous version was releas...

Conditioned [corrected] stimulus informativeness governs conditioned stimulus-unconditioned stimulus associability.

PubMed

Ward, Ryan D; Gallistel, C R; Jensen, Greg; Richards, Vanessa L; Fairhurst, Stephen; Balsam, Peter D

2012-07-01

In a conditioning protocol, the onset of the conditioned stimulus ([CS]) provides information about when to expect reinforcement (unconditioned stimulus [US]). There are two sources of information from the CS in a delay conditioning paradigm in which the CS-US interval is fixed. The first depends on the informativeness, the degree to which CS onset reduces the average expected time to onset of the next US. The second depends only on how precisely a subject can represent a fixed-duration interval (the temporal Weber fraction). In three experiments with mice, we tested the differential impact of these two sources of information on rate of acquisition of conditioned responding (CS-US associability). In Experiment 1, we showed that associability (the inverse of trials to acquisition) increased in proportion to informativeness. In Experiment 2, we showed that fixing the duration of the US-US interval or the CS-US interval or both had no effect on associability. In Experiment 3, we equated the increase in information produced by varying the C/T ratio with the increase produced by fixing the duration of the CS-US interval. Associability increased with increased informativeness, but, as in Experiment 2, fixing the CS-US duration had no effect on associability. These results are consistent with the view that CS-US associability depends on the increased rate of reward signaled by CS onset. The results also provide further evidence that conditioned responding is temporally controlled when it emerges.

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

Effects of weather conditions, light conditions, and road lighting on vehicle speed.

PubMed

Jägerbrand, Annika K; Sjöbergh, Jonas

2016-01-01

Light conditions are known to affect the number of vehicle accidents and fatalities but the relationship between light conditions and vehicle speed is not fully understood. This study examined whether vehicle speed on roads is higher in daylight and under road lighting than in darkness, and determined the combined effects of light conditions, posted speed limit and weather conditions on driving speed. The vehicle speed of passenger cars in different light conditions (daylight, twilight, darkness, artificial light) and different weather conditions (clear weather, rain, snow) was determined using traffic and weather data collected on an hourly basis for approximately 2 years (1 September 2012-31 May 2014) at 25 locations in Sweden (17 with road lighting and eight without). In total, the data included almost 60 million vehicle passes. The data were cleaned by removing June, July, and August, which have different traffic patterns than the rest of the year. Only data from the periods 10:00 A.M.-04:00 P.M. and 06:00 P.M.-10:00 P.M. were used, to remove traffic during rush hour and at night. Multivariate adaptive regression splines was used to evaluate the overall influence of independent variables on vehicle speed and nonparametric statistical testing was applied to test for speed differences between dark-daylight, dark-twilight, and twilight-daylight, on roads with and without road lighting. The results show that vehicle speed in general depends on several independent variables. Analyses of vehicle speed and speed differences between daylight, twilight and darkness, with and without road lighting, did not reveal any differences attributable to light conditions. However, vehicle speed decreased due to rain or snow and the decrease was higher on roads without road lighting than on roads with lighting. These results suggest that the strong association between traffic accidents and darkness or low light conditions could be explained by drivers failing to adjust their

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

Survey of Condition Indicators for Condition Monitoring Systems (Open Access)

DTIC Science & Technology

2014-09-29

Hinesburg, Vermont, 05461, USA jz@renewablenrgsystems.com ABSTRACT Currently, the wind energy industry is swiftly changing its maintenance strategy...from schedule based maintenance to predictive based maintenance . Condition monitoring systems (CMS) play an important role in the predictive... maintenance cycle. As condition monitoring systems are being adopted by more and more OEM and O&M service providers from the wind energy industry, it is

Cladding burst behavior of Fe-based alloys under LOCA

DOE PAGES

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; ...

2015-12-17

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. Themore » most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.« 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

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.

10 CFR 950.11 - Terms and conditions of the Conditional Agreement.

Code of Federal Regulations, 2011 CFR

2011-01-01

... fulfill the conditions precedent specified in § 950.12, subject to certain funding requirements and... the coverage to either the Program Account or the Grant Account. (c) Funding. Each Conditional... anticipated percentage of the total funding in the Program Account to be contributed by appropriated funds to...

10 CFR 950.11 - Terms and conditions of the Conditional Agreement.

Code of Federal Regulations, 2012 CFR

2012-01-01

... fulfill the conditions precedent specified in § 950.12, subject to certain funding requirements and... the coverage to either the Program Account or the Grant Account. (c) Funding. Each Conditional... anticipated percentage of the total funding in the Program Account to be contributed by appropriated funds to...

Fear Conditioning Increases NREM Sleep

PubMed Central

Hellman, Kevin; Abel, Ted

2010-01-01

To understand the role that sleep may play in memory storage, the authors investigated how fear conditioning affects sleep–wake states by performing electroencephalographic (EEG) and electromyographic recordings of C57BL/6J mice receiving fear conditioning, exposure to conditioning stimuli, or immediate shock treatment. This experimental design allowed us to examine the effects of associative learning, presentation of the conditioning stimuli, and presentation of the unconditioned stimuli on sleep–wake states. During the 24 hr after training, fear-conditioned mice had approximately 1 hr more of nonrapid-eye-movement (NREM) sleep and less wakefulness than mice receiving exposure to conditioning stimuli or immediate shock treatment. Mice receiving conditioning stimuli had more delta power during NREM sleep, whereas mice receiving fear conditioning had less theta power during rapid-eye-movement sleep. These results demonstrate that a single trial of fear conditioning alters sleep–wake states and EEG oscillations over a 24-hr period, supporting the idea that sleep is modified by experience and that such changes in sleep–wake states and EEG oscillations may play a role in memory consolidation. PMID:17469920

Watershed condition [Chapter 4

Treesearch

Daniel G. Neary; Jonathan W. Long; Malchus B. Baker

2012-01-01

Managers of the Prescott National Forest are obliged to evaluate the conditions of watersheds under their jurisdiction in order to guide informed decisions concerning grazing allotments, forest and woodland management, restoration treatments, and other management initiatives. Watershed condition has been delineated by contrasts between “good” and “poor” conditions (...

Conditional E-Cash

NASA Astrophysics Data System (ADS)

Shi, Larry; Carbunar, Bogdan; Sion, Radu

We introduce a novel conditional e-cash protocol allowing future anonymous cashing of bank-issued e-money only upon the satisfaction of an agreed-upon public condition. Payers are able to remunerate payees for services that depend on future, yet to be determined outcomes of events. Once payment complete, any double-spending attempt by the payer will reveal its identity; no double-spending by the payee is possible. Payers can not be linked to payees or to ongoing or past transactions. The flow of cash within the system is thus both correct and anonymous. We discuss several applications of conditional e-cash including online trading of financial securities, prediction markets, and betting systems.

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.

Development of a National-Scale Indicator of Benthic Condition for the National Coastal Condition Assessment.

EPA Science Inventory

The US EPA has evaluated the application of a national-scale indicator of estuarine benthic condition for the National Coastal Condition Assessment (NCCA). Historically, in the National Coastal Condition Reports (NCCR I-IV), estuarine benthic condition was assessed by applying m...

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

Interoceptive fear conditioning and panic disorder: the role of conditioned stimulus-unconditioned stimulus predictability.

PubMed

Acheson, Dean T; Forsyth, John P; Moses, Erica

2012-03-01

Interoceptive fear conditioning is at the core of contemporary behavioral accounts of panic disorder. Yet, to date only one study has attempted to evaluate interoceptive fear conditioning in humans (see Acheson, Forsyth, Prenoveau, & Bouton, 2007). That study used brief (physiologically inert) and longer-duration (panicogenic) inhalations of 20% CO(2)-enriched air as an interoceptive conditioned (CS) and unconditioned (US) stimulus and evaluated fear learning in three conditions: CS only, CS-US paired, and CS-US unpaired. Results showed fear conditioning in the paired condition, and fearful responding and resistance to extinction in an unpaired condition. The authors speculated that such effects may be due to difficulty discriminating between the CS and the US. The aims of the present study are to (a) replicate and expand this line of work using an improved methodology, and (b) clarify the role of CS-US discrimination difficulties in either potentiating or depotentiating fear learning. Healthy participants (N=104) were randomly assigned to one of four conditions: (a) CS only, (b) contingent CS-US pairings, (c) unpaired CS and US presentations, or (d) an unpaired "discrimination" contingency, which included an exteroceptive discrimination cue concurrently with CS onset. Electrodermal and self-report ratings served as indices of conditioned responding. Consistent with expectation, the paired contingency and unpaired contingencies yielded elevated fearful responding to the CS alone. Moreover, adding a discrimination cue to the unpaired contingency effectively attenuated fearful responding. Overall, findings are consistent with modern learning theory accounts of panic and highlight the role of interoceptive conditioning and unpredictability in the etiology of panic disorder. Copyright © 2011. Published by Elsevier Ltd.

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

Social Fear Conditioning Paradigm in Virtual Reality: Social vs. Electrical Aversive Conditioning.

PubMed

Reichenberger, Jonas; Porsch, Sonja; Wittmann, Jasmin; Zimmermann, Verena; Shiban, Youssef

2017-01-01

In a previous study we could show that social fear can be induced and extinguished using virtual reality (VR). In the present study, we aimed to investigate the belongingness effect in an operant social fear conditioning (SFC) paradigm which consisted of an acquisition and an extinction phase. Forty-three participants used a joystick to approach different virtual male agents that served as conditioned stimuli. Participants were randomly allocated to one of two experimental conditions. In the electroshock condition, the unconditioned stimulus (US) used during acquisition was an electric stimulation. In the social threat condition, the US consisted of an offense: a spit in the face, mimicked by a sound and a weak air blast to the participant's neck combined with an insult. In both groups the US was presented when participants were close to the agent (75% contingency for CS+). Outcome variables included subjective, psychophysiological and behavioral data. As expected, fear and contingency ratings increased significantly during acquisition and the differentiation between CS+ and CS- vanished during extinction. Furthermore, a clear difference in skin conductance between CS+ and CS- at the beginning of the acquisition indicated that SFC had been successful. However, a fast habituation to the US was found toward the end of the acquisition phase for the physiological response. Furthermore, participants showed avoidance behavior toward CS+ in both conditions. The results show that social fear can successfully be induced and extinguished in VR in a human sample. Thus, our paradigm can help to gain insight into learning and unlearning of social fear. Regarding the belongingness effect, the social threat condition benefits from a better differentiation between the aversive and the non-aversive stimuli. As next step we suggest comparing social-phobic patients to healthy controls in order to investigate possible differences in discrimination learning and to foster the

Social Fear Conditioning Paradigm in Virtual Reality: Social vs. Electrical Aversive Conditioning

PubMed Central

Reichenberger, Jonas; Porsch, Sonja; Wittmann, Jasmin; Zimmermann, Verena; Shiban, Youssef

2017-01-01

In a previous study we could show that social fear can be induced and extinguished using virtual reality (VR). In the present study, we aimed to investigate the belongingness effect in an operant social fear conditioning (SFC) paradigm which consisted of an acquisition and an extinction phase. Forty-three participants used a joystick to approach different virtual male agents that served as conditioned stimuli. Participants were randomly allocated to one of two experimental conditions. In the electroshock condition, the unconditioned stimulus (US) used during acquisition was an electric stimulation. In the social threat condition, the US consisted of an offense: a spit in the face, mimicked by a sound and a weak air blast to the participant’s neck combined with an insult. In both groups the US was presented when participants were close to the agent (75% contingency for CS+). Outcome variables included subjective, psychophysiological and behavioral data. As expected, fear and contingency ratings increased significantly during acquisition and the differentiation between CS+ and CS- vanished during extinction. Furthermore, a clear difference in skin conductance between CS+ and CS- at the beginning of the acquisition indicated that SFC had been successful. However, a fast habituation to the US was found toward the end of the acquisition phase for the physiological response. Furthermore, participants showed avoidance behavior toward CS+ in both conditions. The results show that social fear can successfully be induced and extinguished in VR in a human sample. Thus, our paradigm can help to gain insight into learning and unlearning of social fear. Regarding the belongingness effect, the social threat condition benefits from a better differentiation between the aversive and the non-aversive stimuli. As next step we suggest comparing social-phobic patients to healthy controls in order to investigate possible differences in discrimination learning and to foster the

Some relations between classically conditioned aggression and conditioned suppression in squirrel monkeys.

PubMed Central

Hake, D F; Campbell, R L

1980-01-01

During three experiments with squirrel monkeys, stimulus and shock pairings were given in the presence of a bite tube. Experiments 1 and 2 used a conditioned-suppression procedure in which bar pressing was reinforced with food. A transparent shield prevented biting of the bar. When the stimulus was paired with shock, bar pressing decreased (conditioned suppression) and tube biting increased during the stimulus (classically conditioned aggression). When the bite tube was removed on alternate sessions in Experiment 2, there was more suppression when the tube was present, thus suggesting that biting competed with bar pressing. However, this simple competing-response interpretation was complicated by the findings of Experiment 3 where, with naive monkeys, bar pressing was never reinforced with food, yet bar pressing was induced during the stimulus and was highest when the bite tube was absent. The fact that stimulus-induced bar pressing developed inciated that bar pressing in conditioned-suppression procedures, suppressed or not, may be maintained by two types of control--the food reinforcer and induced CS control. The higher rate of induced bar pressing during the stimulus with the bite tube absent confounds a simple competing response interpretation of conditioned suppression. It suggests that shock-induced responses during conditioned suppression could be both contributing to and competing with responding maintained by food, with the net effect depending on specific but ill-defined features of the situation. PMID:7190996

Emergence of dormant conditioned incentive approach by conditioned withdrawal in nicotine addiction.

PubMed

Scott, Daniel; Hiroi, Noboru

2010-10-15

Nicotine is one of the determinants for the development of persistent smoking, and this maladaptive behavior is characterized by many symptoms, including withdrawal and nicotine seeking. The process by which withdrawal affects nicotine seeking is poorly understood. The impact of a withdrawal-associated cue on nicotine (.2 mg/kg)-conditioned place preference was assessed in male C57BL/6J mice (n = 8-17/group). To establish a cue selectively associated with withdrawal distinct from those associated with nicotine, a tone was paired with withdrawal in their home cages; mice were chronically exposed to nicotine (200 μg/mL for 15 days) from drinking water in their home cages and received the nicotinic acetylcholine receptor antagonist mecamylamine (2.5 mg/kg) to precipitate withdrawal in the presence of a tone. The effect of the withdrawal-associated tone on nicotine-conditioned place preference was then evaluated in the place-conditioning apparatus after a delay, when nicotine-conditioned place preference spontaneously disappeared. A cue associated with precipitated withdrawal reactivated the dormant effect of nicotine-associated cues on conditioned place preference. This effect occurred during continuous exposure to nicotine but not during abstinence. A conditioned withdrawal cue could directly amplify the incentive properties of cues associated with nicotine. This observation extends the contemporary incentive account of the role of withdrawal in addiction to cue-cue interaction. Copyright © 2010 Society of Biological Psychiatry. Published by Elsevier Inc. All rights reserved.

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

Faidy, C.; Gilles, P.

The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of:more » simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.« less

Hot conditioning equipment conceptual design report

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

Bradshaw, F.W., Westinghouse Hanford

1996-08-06

This report documents the conceptual design of the Hot Conditioning System Equipment. The Hot conditioning System will consist of two separate designs: the Hot Conditioning System Equipment; and the Hot Conditioning System Annex. The Hot Conditioning System Equipment Design includes the equipment such as ovens, vacuum pumps, inert gas delivery systems, etc.necessary to condition spent nuclear fuel currently in storage in the K Basins of the Hanford Site. The Hot Conditioning System Annex consists of the facility of house the Hot Conditioning System. The Hot Conditioning System will be housed in an annex to the Canister Storage Building. The Hotmore » Conditioning System will consist of pits in the floor which contain ovens in which the spent nuclear will be conditioned prior to interim storage.« less

The emotional and academic consequences of parental conditional regard: comparing conditional positive regard, conditional negative regard, and autonomy support as parenting practices.

PubMed

Roth, Guy; Assor, Avi; Niemiec, Christopher P; Deci, Edward L; Ryan, Richard M

2009-07-01

The authors conducted 2 studies of 9th-grade Israeli adolescents (169 in Study 1, 156 in Study 2) to compare the parenting practices of conditional positive regard, conditional negative regard, and autonomy support using data from multiple reporters. Two socialization domains were studied: emotion control and academics. Results were consistent with the self-determination theory model of internalization, which posits that (a) conditional negative regard predicts feelings of resentment toward parents, which then predict dysregulation of negative emotions and academic disengagement; (b) conditional positive regard predicts feelings of internal compulsion, which then predict suppressive regulation of negative emotions and grade-focused academic engagement; and (c) autonomy support predicts sense of choice, which then predicts integrated regulation of negative emotions and interest-focused academic engagement. These findings suggest that even parents' use of conditional positive regard as a socialization practice has adverse emotional and academic consequences, relative to autonomy support.

Condition sensor system and method

NASA Technical Reports Server (NTRS)

Polhemus, J. T.; Morgan, J. E.; Mandell, A. (Inventor)

1978-01-01

The condition sensor system comprises a condition detector which produces a pulse when a parameter of the monitored condition exceeds a desired threshold. A resettable condition counter counts each pulse. A resettable timer is preset to produce a particular time frame. The counter produces a condition signal when the accumulated number of pulses within the time frame is equal to or greater than a preset count. Control means responsive to the incoming pulses and to the condition signal produce control signals that control utilization devices. After a suitable delay, the last detected pulse simultaneously resets the pulse counter and the timer, and prepares them for sensing another condition occurrence within the time frame. The invention has particular utility in the process of detecting rocking motions of blind people. A controlled, audible, bio-feedback signal is provided which constitutes a warning to the blind person that he is rocking.

The Association of Skin Conditions with Housing Conditions Among North Carolina Latino Migrant Farmworkers

PubMed Central

Gustafson, Cheryl J.; Feldman, Steven R.; Quandt, Sara A.; Isom, Scott; Chen, Haiying; Spears, Chaya R.; Arcury, Thomas A.

2012-01-01

Background Skin conditions are common among Latino migrant farmworkers. Although many skin conditions are related to occupational exposures, poor housing conditions may also contribute to skin ailments in migrant farmworkers. Objectives To evaluate the association between housing conditions and skin conditions among Latino migrant farmworkers. Methods A cross-sectional study design using interview questionnaires, home inspections, and environmental sampling was implemented to document housing quality of farmworker camps/homes, and the prevalence of self-reported skin conditions in Latino migrant farmworkers. Interviews were completed with 371 farmworkers residing in 186 of the 226 camps (camp response rate 82.3%). Results Self-reported pruritus (31%), rash (25%), scaling (12%), blisters (11%), and ingrown nails (10%) were commonly among the participants. Pruritus was more likely to be reported by farmworkers living in dwellings without air conditioning (p<0.05). Rash was associated with dwellings reported to have a low humidity (p<0.05). Scaling was more likely to be reported by farmworkers living in dwellings with indoor temperatures in the thermal discomfort range (p<0.05). No statistically significant associations were detected for indoor allergens and self-reported skin ailments among migrant farmworkers. Conclusions Skin conditions are common among migrant farmworkers in North Carolina. The quality of housing conditions, particularly hot, dry indoor thermal environment, demonstrated significant associations with pruritus, rash, and scaling. The impact of housing characteristics on pruritus and blisters was greatest in new migrant farmworkers. Further research is needed to delineate additional housing factors that could cause or exacerbate skin diseases in farmworkers. PMID:23675774

A method for simultaneously counterbalancing condition order and assignment of stimulus materials to conditions.

PubMed

Zeelenberg, René; Pecher, Diane

2015-03-01

Counterbalanced designs are frequently used in the behavioral sciences. Studies often counterbalance either the order in which conditions are presented in the experiment or the assignment of stimulus materials to conditions. Occasionally, researchers need to simultaneously counterbalance both condition order and stimulus assignment to conditions. Lewis (1989; Behavior Research Methods, Instruments, & Computers 25:414-415, 1993) presented a method for constructing Latin squares that fulfill these requirements. The resulting Latin squares counterbalance immediate sequential effects, but not remote sequential effects. Here, we present a new method for generating Latin squares that simultaneously counterbalance both immediate and remote sequential effects and assignment of stimuli to conditions. An Appendix is provided to facilitate implementation of these Latin square designs.

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

Pavlovian second-order conditioned analgesia.

PubMed

Ross, R T

1986-01-01

Three experiments with rat subjects assessed conditioned analgesia in a Pavlovian second-order conditioning procedure by using inhibition of responding to thermal stimulation as an index of pain sensitivity. In Experiment 1, rats receiving second-order conditioning showed longer response latencies during a test of pain sensitivity in the presence of the second-order conditioned stimulus (CS) than rats receiving appropriate control procedures. Experiment 2 found that extinction of the first-order CS had no effect on established second-order conditioned analgesia. Experiment 3 evaluated the effects of post second-order conditioning pairings of morphine and the shock unconditioned stimulus (US). Rats receiving paired morphine-shock presentations showed significantly shorter response latencies during a hot-plate test of pain sensitivity in the presence of the second-order CS than did groups of rats receiving various control procedures; second-order analgesia was attenuated. These data extend the associative account of conditioned analgesia to second-order conditioning situations and are discussed in terms of the mediation of both first- and second-order analgesia by an association between the CS and a representation or expectancy of the US, which may directly activate endogenous pain inhibition systems.

A combination of body condition measurements is more informative than conventional condition indices: temporal variation in body condition and corticosterone in brown tree snakes (Boiga irregularis).

PubMed

Waye, Heather L; Mason, Robert T

2008-02-01

The body condition index is a common method for quantifying the energy reserves of individual animals. Because good body condition is necessary for reproduction in many species, body condition indices can indicate the potential reproductive output of a population. Body condition is related to glucocorticoid production, in that low body condition is correlated to high concentrations of corticosterone in reptiles. We compared the body condition index and plasma corticosterone levels of brown tree snakes on Guam in 2003 to those collected in 1992/1993 to determine whether that population still showed the chronic stress and poor condition apparent in the earlier study. We also examined the relationship between fat mass, body condition and plasma corticosterone concentrations as indicators of physiological condition of individuals in the population. Body condition was significantly higher in 2003 than in the earlier sample for mature male and female snakes, but not for juveniles. The significantly lower levels of corticosterone in all three groups in 2003 suggests that although juveniles did not have significantly improved energy stores they, along with the mature males and females, were no longer under chronic levels of stress. Although the wet season of 2002 was unusually rainy, low baseline levels of corticosterone measured in 2000 indicate that the improved body condition of snakes in 2003 is likely the result of long-term changes in prey populations rather than annual variation in response to environmental conditions.

Causal conditionals and counterfactuals

PubMed Central

Frosch, Caren A.; Byrne, Ruth M.J.

2012-01-01

Causal counterfactuals e.g., ‘if the ignition key had been turned then the car would have started’ and causal conditionals e.g., ‘if the ignition key was turned then the car started’ are understood by thinking about multiple possibilities of different sorts, as shown in six experiments using converging evidence from three different types of measures. Experiments 1a and 1b showed that conditionals that comprise enabling causes, e.g., ‘if the ignition key was turned then the car started’ primed people to read quickly conjunctions referring to the possibility of the enabler occurring without the outcome, e.g., ‘the ignition key was turned and the car did not start’. Experiments 2a and 2b showed that people paraphrased causal conditionals by using causal or temporal connectives (because, when), whereas they paraphrased causal counterfactuals by using subjunctive constructions (had…would have). Experiments 3a and 3b showed that people made different inferences from counterfactuals presented with enabling conditions compared to none. The implications of the results for alternative theories of conditionals are discussed. PMID:22858874

Conditional Belief Types

DTIC Science & Technology

2016-04-19

event is the same as conditioning on the event being certain, which formalizes the standard informal interpretation of conditional probability. The game ...theoretic application of our model, discussed within an example, sheds light on a number of issues in the analysis of extensive form games . Type...belief types Block 13: Supplementary Note © 2014 . Published in Games and Economic Behavior, Vol. Ed. 0 87, (0) (2014), (, (0). DoD Components

How Are Genetic Conditions Diagnosed?

MedlinePlus

... Consultation How are genetic conditions diagnosed? How are genetic conditions diagnosed? A doctor may suspect a diagnosis ... and advocacy resources. For more information about diagnosing genetic conditions: Genetics Home Reference provides information about genetic ...

Dissociating Contingency Awareness and Conditioned Attitudes: Evidence of Contingency-Unaware Evaluative Conditioning

ERIC Educational Resources Information Center

Hutter, Mandy; Sweldens, Steven; Stahl, Christoph; Unkelbach, Christian; Klauer, Karl Christoph

2012-01-01

Whether human evaluative conditioning can occur without contingency awareness has been the subject of an intense and ongoing debate for decades, troubled by a wide array of methodological difficulties. Following recent methodological innovations, the available evidence currently points to the conclusion that evaluative conditioning effects do not…

Teachers' Working Conditions. Findings from "The Condition of Education, 1996," No. 7.

ERIC Educational Resources Information Center

Choy, Susan P.

Working conditions play an important role in a school's ability to attract, develop, and retain effective teachers. Data presented here describe a number of aspects of teachers' working conditions, including workload, compensation, school and district support for teachers' professional development, school decision making, school safety, student…

Interpersonal Attraction Under Negative Conditions: A Problem for the Classical Conditioning Paradigm?

ERIC Educational Resources Information Center

Kenrick, Douglas T.; Johnson, Gregory A.

The influence of aversive conditions on interpersonal attraction was investigated using 60 female undergraduates as subjects. Dyads were formed and equally divided into aversive (loud-noise) and neutral (low-noise) conditions. After completing an attitude survey questionnaire subjects completed a short filler task while the experimenter…

National Coastal Condition Assessment

EPA Pesticide Factsheets

It is important to monitor coastal waters for potentially harmful trends and to identify areas in good condition. That is the purpose of the National Coastal Condition Assessment, which EPA conducts every few years.

Teaching and Demonstrating Classical Conditioning.

ERIC Educational Resources Information Center

Sparrow, John; Fernald, Peter

1989-01-01

Discusses classroom demonstrations of classical conditioning and notes tendencies to misrepresent Pavlov's procedures. Describes the design and construction of the conditioner that is used for demonstrating classical conditioning. Relates how students experience conditioning, generalization, extinction, discrimination, and spontaneous recovery.…

A Ranking-Theoretic Approach to Conditionals

ERIC Educational Resources Information Center

Spohn, Wolfgang

2013-01-01

Conditionals somehow express conditional beliefs. However, conditional belief is a bi-propositional attitude that is generally not truth-evaluable, in contrast to unconditional belief. Therefore, this article opts for an expressivistic semantics for conditionals, grounds this semantics in the arguably most adequate account of conditional belief,…

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

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.

76 FR 14600 - Dental Conditions

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-17

... DEPARTMENT OF VETERANS AFFAIRS 38 CFR Part 3 RIN 2900-AN28 Dental Conditions AGENCY: Department of... its adjudication regulations regarding service connection of dental conditions for treatment purposes... Administration (VBA) of service connection of dental conditions for the purpose of establishing eligibility for...

Breast Changes and Conditions

Cancer.gov

Breast changes include benign conditions and those that increase the risk of breast cancer. Symptoms and treatment of breast conditions such as ADH, ALH, cysts, DCIS, and LCIS are explained to women who may have received an abnormal mammogram finding.

Compatibility Conditions of Structural Mechanics

NASA Technical Reports Server (NTRS)

Patnaik, Surya N.; Coroneos, Rula M.; Hopkins, Dale A.

1999-01-01

The theory of elasticity has camouflaged a deficiency in the compatibility formulation since 1860. In structures the ad hoc compatibility conditions through virtual "cuts" and closing "gaps" are not parallel to the strain formulation in elasticity. This deficiency in the compatibility conditions has prevented the development of a direct stress determination method in structures and in elasticity. We have addressed this deficiency and attempted to unify the theory of compatibility. This work has led to the development of the integrated force method for structures and the completed Beltrami-Michell formulation for elasticity. The improved accuracy observed in the solution of numerical examples by the integrated force method can be attributed to the compliance of the compatibility conditions. Using the compatibility conditions allows mapping of variables and facile movement among different structural analysis formulations. This paper reviews and illustrates the requirement of compatibility in structures and in elasticity. It also describes the generation of the conditions and quantifies the benefits of their use. The traditional analysis methods and available solutions (which have been obtained bypassing the missed conditions) should be verified for compliance of the compatibility conditions.

Teacher Working Conditions that Matter

ERIC Educational Resources Information Center

Leithwood, Ken; McAdie, Pat

2007-01-01

To advance understanding of the issues concerning teachers' working conditions, the Elementary Teachers' Federation of Ontario commissioned one of the authors to do an analytical review of literature on teachers' working conditions. This resulted in the publication, "Teacher Working Conditions That Matter: Evidence for Change." The…

Contractual conditions, working conditions and their impact on health and well-being.

PubMed

Robone, Silvana; Jones, Andrew M; Rice, Nigel

2011-10-01

Given changes in the labour market in past decades, it is of interest to evaluate whether and how contractual and working conditions affect health and psychological well-being in society today. We consider the effects of contractual and working conditions on self-assessed health and psychological well-being using twelve waves (1991/1992-2002/2003) of the British Household Panel Survey. For self-assessed health, the dependent variable is categorical, and we estimate non-linear dynamic panel ordered probit models, while for psychological well-being, we estimate a dynamic linear specification. The results show that both contractual and working conditions have an influence on health and psychological well-being and that the impact is different for men and women.

Rapid condition assessment of structural condition after a blast using state-space identification

NASA Astrophysics Data System (ADS)

Eskew, Edward; Jang, Shinae

2015-04-01

After a blast event, it is important to quickly quantify the structural damage for emergency operations. In order improve the speed, accuracy, and efficiency of condition assessments after a blast, the authors have previously performed work to develop a methodology for rapid assessment of the structural condition of a building after a blast. The method involved determining a post-event equivalent stiffness matrix using vibration measurements and a finite element (FE) model. A structural model was built for the damaged structure based on the equivalent stiffness, and inter-story drifts from the blast are determined using numerical simulations, with forces determined from the blast parameters. The inter-story drifts are then compared to blast design conditions to assess the structures damage. This method still involved engineering judgment in terms of determining significant frequencies, which can lead to error, especially with noisy measurements. In an effort to improve accuracy and automate the process, this paper will look into a similar method of rapid condition assessment using subspace state-space identification. The accuracy of the method will be tested using a benchmark structural model, as well as experimental testing. The blast damage assessments will be validated using pressure-impulse (P-I) diagrams, which present the condition limits across blast parameters. Comparisons between P-I diagrams generated using the true system parameters and equivalent parameters will show the accuracy of the rapid condition based blast assessments.

Teacher Working Conditions are Student Learning Conditions. A Report to Mike Easley on the 2004 North Carolina Teacher Working Conditions Survey

ERIC Educational Resources Information Center

Southeast Center for Teaching Quality (The), University of North Carolina, 2004

2004-01-01

For virtually any business or organization, the conditions in which employees work drive their satisfaction and productivity. Yet, while business often focuses on employee satisfaction, many schools often struggle to address critical working conditions -- isolating teachers in classrooms with closed doors, denying them basic materials to do their…

Operant Conditioning for Special Educators.

ERIC Educational Resources Information Center

Pedrini, Bonnie C.; Pedrini, D. T.

The paper briefly explains operant conditioning as it pertains to special educators. Operant conditioning is thought to be an efficient method for modifying student behavior. Using the B. F. Skinner frame of reference, operant conditioning is said to include behavior modification and therapy, programed instruction, and computer assisted and…

The Pavlovian analysis of instrumental conditioning.

PubMed

Gormezano, I; Tait, R W

1976-01-01

An account was given of the development within the Russian literature of a uniprocess formulation of classical and instrumental conditioning, known as the bidirectional conditioning hypothesis. The hypothesis purports to offer a single set of Pavlovian principles to account for both paradigms, based upon a neural model which assumes that bidirectional (forward and backward) connections are formed in both calssical and instrumental conditioning situations. In instrumental conditioning, the bidirectional connections are hypothesized to be simply more complex than those in classical conditioning, and any differences in empirical functions are presumed to lie not in difference in mechanism, but in the strength of the forward and backward connections. Although bidirectional connections are assumed to develop in instrumental conditioning, the experimental investigation of the bidirectional conditioning hypothesis has been essentially restricted to the classical conditioning operations of pairing two CSs (sensory preconditioning training), a US followed by a CS (backward conditioning training) and two USs. However, the paradigm involving the pairing of two USs, because of theoretical and analytical considerations, is the one most commonly employed by Russian investigators. The results of an initial experiment involving the pairing of two USs, and reference to the results of a more extensive investigation, leads us to tentatively question the validity of the bidirectional conditioning account of instrumental conditioning.

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.

Shell Condition and Survival of Puget Sound Pteropods Are Impaired by Ocean Acidification Conditions

PubMed Central

Busch, D. Shallin; Maher, Michael; Thibodeau, Patricia; McElhany, Paul

2014-01-01

We tested whether the thecosome pteropod Limacina helicina from Puget Sound, an urbanized estuary in the northwest continental US, experiences shell dissolution and altered mortality rates when exposed to the high CO2, low aragonite saturation state (Ωa) conditions that occur in Puget Sound and the northeast Pacific Ocean. Five, week-long experiments were conducted in which we incubated pteropods collected from Puget Sound in four carbon chemistry conditions: current summer surface (∼460–500 µatm CO2, Ωa≈1.59), current deep water or surface conditions during upwelling (∼760 and ∼1600–1700 µatm CO2, Ωa≈1.17 and 0.56), and future deep water or surface conditions during upwelling (∼2800–3400 µatm CO2, Ωa≈0.28). We measured shell condition using a scoring regime of five shell characteristics that capture different aspects of shell dissolution. We characterized carbon chemistry conditions in statistical analyses with Ωa, and conducted analyses considering Ωa both as a continuous dataset and as discrete treatments. Shell dissolution increased linearly as aragonite saturation state decreased. Discrete treatment comparisons indicate that shell dissolution was greater in undersaturated treatments compared to oversaturated treatments. Survival increased linearly with aragonite saturation state, though discrete treatment comparisons indicated that survival was similar in all but the lowest saturation state treatment. These results indicate that, under starvation conditions, pteropod survival may not be greatly affected by current and expected near-future aragonite saturation state in the NE Pacific, but shell dissolution may. Given that subsurface waters in Puget Sound’s main basin are undersaturated with respect to aragonite in the winter and can be undersaturated in the summer, the condition and persistence of the species in this estuary warrants further study. PMID:25162395

Shell condition and survival of Puget Sound pteropods are impaired by ocean acidification conditions.

PubMed

Busch, D Shallin; Maher, Michael; Thibodeau, Patricia; McElhany, Paul

2014-01-01

We tested whether the thecosome pteropod Limacina helicina from Puget Sound, an urbanized estuary in the northwest continental US, experiences shell dissolution and altered mortality rates when exposed to the high CO2, low aragonite saturation state (Ωa) conditions that occur in Puget Sound and the northeast Pacific Ocean. Five, week-long experiments were conducted in which we incubated pteropods collected from Puget Sound in four carbon chemistry conditions: current summer surface (∼460-500 µatm CO2, Ωa≈1.59), current deep water or surface conditions during upwelling (∼760 and ∼1600-1700 µatm CO2, Ωa≈1.17 and 0.56), and future deep water or surface conditions during upwelling (∼2800-3400 µatm CO2, Ωa≈0.28). We measured shell condition using a scoring regime of five shell characteristics that capture different aspects of shell dissolution. We characterized carbon chemistry conditions in statistical analyses with Ωa, and conducted analyses considering Ωa both as a continuous dataset and as discrete treatments. Shell dissolution increased linearly as aragonite saturation state decreased. Discrete treatment comparisons indicate that shell dissolution was greater in undersaturated treatments compared to oversaturated treatments. Survival increased linearly with aragonite saturation state, though discrete treatment comparisons indicated that survival was similar in all but the lowest saturation state treatment. These results indicate that, under starvation conditions, pteropod survival may not be greatly affected by current and expected near-future aragonite saturation state in the NE Pacific, but shell dissolution may. Given that subsurface waters in Puget Sound's main basin are undersaturated with respect to aragonite in the winter and can be undersaturated in the summer, the condition and persistence of the species in this estuary warrants further study.

SETTING EXPECTATIONS FOR THE ECOLOGICAL CONDITION OF STREAMS: THE CONCEPT OF REFERENCE CONDITION

EPA Science Inventory

An important component of the biological assessment of stream condition is an evaluation of the direct or indirect effects of human activities or disturbances. The concept of a "reference condition" is increasingly used to describe the standard or benchmark against which current ...

Dual Effects on Choice of Conditioned Reinforcement Frequency and Conditioned Reinforcement Value

ERIC Educational Resources Information Center

McDevitt, Margaret A.; Williams, Ben A.

2010-01-01

Pigeons were presented with a concurrent-chains schedule in which the total time to primary reinforcement was equated for the two alternatives (VI 30 s VI 60 s vs. VI 60 s VI 30 s). In one set of conditions, the terminal links were signaled by the same stimulus, and in another set of conditions they were signaled by different stimuli. Choice was…

The causal structure of utility conditionals.

PubMed

Bonnefon, Jean-François; Sloman, Steven A

2013-01-01

The psychology of reasoning is increasingly considering agents' values and preferences, achieving greater integration with judgment and decision making, social cognition, and moral reasoning. Some of this research investigates utility conditionals, ''if p then q'' statements where the realization of p or q or both is valued by some agents. Various approaches to utility conditionals share the assumption that reasoners make inferences from utility conditionals based on the comparison between the utility of p and the expected utility of q. This article introduces a new parameter in this analysis, the underlying causal structure of the conditional. Four experiments showed that causal structure moderated utility-informed conditional reasoning. These inferences were strongly invited when the underlying structure of the conditional was causal, and significantly less so when the underlying structure of the conditional was diagnostic. This asymmetry was only observed for conditionals in which the utility of q was clear, and disappeared when the utility of q was unclear. Thus, an adequate account of utility-informed inferences conditional reasoning requires three components: utility, probability, and causal structure. Copyright © 2012 Cognitive Science Society, Inc.

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.

PWR upper/lower internals shield

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

Homyk, W.A.

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use ofmore » lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.« 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.

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.

Pavlovian conditioning with ethanol: sign-tracking (autoshaping), conditioned incentive, and ethanol self-administration.

PubMed

Krank, Marvin D

2003-10-01

Conditioned incentive theories of addictive behavior propose that cues signaling a drug's reinforcing effects activate a central motivational state. Incentive motivation enhances drug-taking and drug-seeking behavior. We investigated the behavioral response to cues associated with ethanol and their interaction with operant self-administration of ethanol. In two experiments, rats received operant training to press a lever for a sweetened ethanol solution. After operant training, the animals were given Pavlovian pairings of a brief and localized cue light with the sweetened ethanol solution (no lever present). Lever pressing for ethanol was then re-established, and the behavioral effects of the cue light were tested during an ethanol self-administration session. The conditioned responses resulting from pairing cue lights with the opportunity to ingest ethanol had three main effects: (1) induction of operant behavior reinforced by ethanol, (2) stimulation of ethanol-seeking behavior (magazine entries), and (3) signal-directed behavior (i.e., autoshaping, or sign-tracking). Signal-directed behavior interacted with the other two effects in a manner predicted by the location of the cue light. These conditioned responses interact with operant responding for ethanol reinforcement. These findings demonstrate the importance of Pavlovian conditioning effects on ethanol self-administration and are consistent with conditioned incentive theories of addictive behavior.

The Effects of Alarm Display, Processing, and Availability on Crew Performance

DTIC Science & Technology

2000-11-01

snow Instrumentation line leakage Small LOCA Steam generator tube rupture Small feedwater leakage inside containment Cycling of main steam...implemented. • Due to primary pressure controller failure, pressure heater banks cycle between on and off. 8.00 CF1 CF2 CF3 CF4 CF5...temperatures after the high-pressure pre- heaters flows into the steam generators number of active emergency feedwater pumps openings of the condensate

Particle Size Determination in Small Solid Propellant Rocket Motors Using the Diffractively Scattered Light Method.

DTIC Science & Technology

1982-10-01

calibrated by using spherical glass beads and aluminum oxide powder . Measurements were successfully made at both locations. Because DO 1473 EoITioN OF I NOVy...determined using measurements of diffrac- tively scattered laser power spectra. The apparatus was calibrated by using spherical glass beads and aluminum oxide... powder . Measurements were successfully made at both loca- tions. Because of the presence of char agglomerates in the exhaust, continued effort is

Atmospheric Models For Over-Ocean Propagation Loss

DTIC Science & Technology

2015-08-24

Radiosonde balloons are launched daily at selected loca- tions, and measure temperature, dew point temperature, and air pressure as they ascend. Radiosondes...different times of year and locations. The result was used to estimate high-reliability SHF/EHF air -to-surface radio link performance in a maritime...environment. I. INTRODUCTION Air -to-surface radio links differ from typical satellite com- munications links in that the path elevation angles are lower

Strategy and Airpower

DTIC Science & Technology

2011-01-01

myopia often leads otherwise competent observers to under­ estimate significantly the new technology’s potential. Two business examples stand out: in...direction. With precision of effect combined with precision of impact, bloodless war becomes a reality . To this point, we have tried to make the...against virtually all of the centers of gravity directly related to strategic objectives, regardless of their loca­ tion. Because it can bring many

Joint Force Quarterly. Number 15, Spring 1997

DTIC Science & Technology

1997-06-01

headquarters to extract information from sensors on the vehicle without bothering crew members with extraneous reports. Position loca- tion devices on... change in how they do business. Air Force lean logistics and Army velocity management programs are literal springboards for quantum improvements in...Spring 1997 Victory smiles upon those who anticipate the changes in the character of war, not upon those who wait to adapt themselves after the changes

Fluid Dynamic Analysis of Volcanic Tremor,

DTIC Science & Technology

1982-10-01

information regarding the fluid system Fiske (1969) Kilauea volcano : The 1967-68 summit configuration, tremor magnitudes and source loca- eruption...Koyanagi (1981) Deep volcanic tremor logicalSociety of America, vol. 40, p. 175-194. and magma ascent mechanism under Kilauea , Hawaii . Omori, F...dynamics Seismology Tremors Volcanoes 40 M\\ TlACT (amhue ai revers if5 neeeeiy md ide~Wify by block number) Low-frequency (< 10 Hz) volcanic earthquakes

Electrical Aversion Conditioning with Chronic Alcoholics

ERIC Educational Resources Information Center

Vogler, Roger E.; And Others

1970-01-01

Pseudoconditioning (random shock delivery), sham conditioning (no shock), and ward controls (routine hospital treatment) ( were compared with two conditioning groups. Conditioning only (contingent shock) and booster Ss (additional conditioning sessions after release from hospital) were shocked for drinking and reinforced by shock termination for…

Post-conditioning experience with acute or chronic inflammatory pain reduces contextual fear conditioning in the rat

PubMed Central

Johnston, Ian N.; Maier, Steven F.; Rudy, Jerry W.; Watkins, Linda R.

2017-01-01

There is evidence that pain can impact cognitive function in people. The present study evaluated whether Pavlovian fear conditioning in rats would be reduced if conditioning were followed by persistent inflammatory pain induced by a subcutaneous injection of dilute formalin or complete Freund's adjuvant (CFA) on the dorsal lumbar surface of the back. Formalin-induced pain specifically impaired contextual fear conditioning but not auditory cue conditioning (Experiment 1A). Moreover, formalin pain only impaired contextual fear conditioning if it was initiated within 1 h of conditioning and did not have a significant effect if initiated 2, 8 or 32 h after (Experiments 1A and 1B). Experiment 2 showed that formalin pain initiated after a session of context pre-exposure reduced the ability of that pre-exposure to facilitate contextual fear when the rat was limited to a brief exposure to the context during conditioning. Similar impairments in context- but not CS-fear conditioning were also observed if the rats received an immediate post-conditioning injection with CFA (Experiment 3). Finally, we confirmed that formalin and CFA injected s.c. on the back induced pain-indicative behaviours, hyperalgesia and allodynia with a similar timecourse to intraplantar injections (Experiment 4). These results suggest that persistent pain impairs learning in a hippocampus-dependent task, and may disrupt processes that encode experiences into long-term memory. PMID:21920390

Disruption of Responding Maintained by Conditioned Reinforcement: Alterations in Response-Conditioned-Reinforcer Relations

ERIC Educational Resources Information Center

Lieving, Gregory A.; Reilly, Mark P.; Lattal, Kennon A.

2006-01-01

An observing procedure was used to investigate the effects of alterations in response-conditioned-reinforcer relations on observing. Pigeons responded to produce schedule-correlated stimuli paired with the availability of food or extinction. The contingency between observing responses and conditioned reinforcement was altered in three experiments.…

The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

PubMed

Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

2009-09-01

In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

Choice and conditioned reinforcement.

PubMed Central

Fantino, E; Freed, D; Preston, R A; Williams, W A

1991-01-01

A potential weakness of one formulation of delay-reduction theory is its failure to include a term for rate of conditioned reinforcement, that is, the rate at which the terminal-link stimuli occur in concurrent-chains schedules. The present studies assessed whether or not rate of conditioned reinforcement has an independent effect upon choice. Pigeons responded on either modified concurrent-chains schedules or on comparable concurrent-tandem schedules. The initial link was shortened on only one of two concurrent-chains schedules and on only one of two corresponding concurrent-tandem schedules. This manipulation increased rate of conditioned reinforcement sharply in the chain but not in the tandem schedule. According to a formulation of delay-reduction theory, when the outcomes chosen (the terminal links) are equal, as in Experiment 1, choice should depend only on rate of primary reinforcement; thus, choice should be equivalent for the tandem and chain schedules despite a large difference in rate of conditioned reinforcement. When the outcomes chosen are unequal, however, as in Experiment 2, choice should depend upon both rate of primary reinforcement and relative signaled delay reduction; thus, larger preferences should occur in the chain than in the tandem schedules. These predictions were confirmed, suggesting that increasing the rate of conditioned reinforcement on concurrent-chains schedules may have no independent effect on choice. PMID:2037826

Conditioned pain modulation is affected by occlusion cuff conditioning stimulus intensity, but not duration.

PubMed

Smith, A; Pedler, A

2018-01-01

Various conditioned pain modulation (CPM) methodologies have been used to investigate diffuse noxious inhibitory control pain mechanisms in healthy and clinical populations. Occlusion cuff parameters have been poorly studied. We aimed to investigate whether occlusion cuff intensity and/or duration influenced CPM magnitudes. We also investigated the role of physical activity levels on CPM magnitude. Two studies were performed to investigate the role of intensity and duration of occlusion cuff conditioning stimulus on test stimulus (tibialis anterior pressure pain thresholds). In Study 1, conditioning stimulus intensity of 2/10 or 5/10 (duration <20 s) was evaluated using a paired-samples t-test. In Study 2, duration of 2/10 conditioning stimulus was 3 min. One-way repeated-measures ANOVA was used to investigate the effect of time (0, 1, 2 and 3 min) on CPM magnitude. In Study 1, 27 healthy volunteers (mean ± SD: 24.9 years (±4.5); eight female) demonstrated that an occlusion cuff applied to the upper arm eliciting 5/10 local pain resulted in a significant (mean ± SD: 17% ± 46%) increase in CPM magnitude, when compared to 2/10 intensity (-3% ± 38%, p = 0.026), whereas in Study 2, 25 healthy volunteers (22.5 years (±2.7); 13 female) demonstrated that 3 min of 2/10 CS intensity did not result in a significant change in CPM (p = 0.21). There was no significant relationship between physical activity levels and CPM in either study (p > 0.22). This study demonstrated that an occlusion cuff of 5/10 conditioning stimulus intensity, when compared to 2/10, significantly increased CPM magnitude. Maintaining 2/10 conditioning stimulus for 3 min did not increase CPM magnitude. Dysfunctional conditioned pain modulation (CPM) has been associated with poor health outcomes. Various factors can influence CPM outcomes. The role of occlusion cuff conditioning stimulus intensity and duration has not been previously investigated. Intensity (5/10), but not

Thermal conditions and perceived air quality in an air-conditioned auditorium

NASA Astrophysics Data System (ADS)

Polednik, Bernard; Guz, Łukasz; Skwarczyński, Mariusz; Dudzińska, Marzenna R.

2016-07-01

The study reports measurements of indoor air temperature (T) and relative humidity (RH), perceived air quality (PAQ) and CO2, fine aerosol particle number (PN) and mass (PM1) concentrations in an air conditioned auditorium. The measurements of these air physical parameters have been carried out in the unoccupied auditorium with the air conditioning system switched off (AC off mode) and in the unoccupied and occupied auditorium with the air conditioning system switched off during the night and switched on during the day (AC on/off mode). The average indoor air thermal parameters, CO2 concentration and the PAQ value (in decipols) were elevated, while average PM1 concentration was lower in the AC on/off mode. A statistically significant (p < 0.001) positive correlation has been observed between T and PAQ values and CO2 concentrations (r = 0.66 and r = 0.59, respectively) in that AC mode. A significant negative correlation has been observed between T and PN and PM1 concentrations (r = -0.38 and r = -0.49, respectively). In the AC off mode the above relations between T and the particle concentrations were not that unequivocal. These findings may be of importance as they indicate that in certain AC operation modes the indoor air quality deteriorates along with the variation of the indoor air microclimate and room occupation. This, in turn, may adversely affect the comfort and productivity of the users of air conditioned premises.

Chapter 7 - Crown condition

Treesearch

KaDonna C. Randolph

2018-01-01

Tree crown conditions are visually assessed by the U.S. Department of Agriculture, Forest Service, Forest Inventory and Analysis (FIA) Program as an indicator of forest health. These assessments are useful because individual tree photosynthetic capacity is dependent upon the size and condition of the crown. In general, trees with full, vigorous crowns are associated...

Medial Auditory Thalamic Stimulation as a Conditioned Stimulus for Eyeblink Conditioning in Rats

ERIC Educational Resources Information Center

Campolattaro, Matthew M.; Halverson, Hunter E.; Freeman, John H.

2007-01-01

The neural pathways that convey conditioned stimulus (CS) information to the cerebellum during eyeblink conditioning have not been fully delineated. It is well established that pontine mossy fiber inputs to the cerebellum convey CS-related stimulation for different sensory modalities (e.g., auditory, visual, tactile). Less is known about the…

The Emotional and Academic Consequences of Parental Conditional Regard: Comparing Conditional Positive Regard, Conditional Negative Regard, and Autonomy Support as Parenting Practices

ERIC Educational Resources Information Center

Roth, Guy; Assor, Avi; Niemiec, Christopher P.; Deci, Edward L.; Ryan, Richard M.

2009-01-01

The authors conducted 2 studies of 9th-grade Israeli adolescents (169 in Study 1, 156 in Study 2) to compare the parenting practices of conditional positive regard, conditional negative regard, and autonomy support using data from multiple reporters. Two socialization domains were studied: emotion control and academics. Results were consistent…

Jump conditions in transonic equilibria

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

Guazzotto, L.; Betti, R.; Jardin, S. C.

2013-04-15

In the present paper, the numerical calculation of transonic equilibria, first introduced with the FLOW code in Guazzotto et al.[Phys. Plasmas 11, 604 (2004)], is critically reviewed. In particular, the necessity and effect of imposing explicit jump conditions at the transonic discontinuity are investigated. It is found that 'standard' (low-{beta}, large aspect ratio) transonic equilibria satisfy the correct jump condition with very good approximation even if the jump condition is not explicitly imposed. On the other hand, it is also found that high-{beta}, low aspect ratio equilibria require the correct jump condition to be explicitly imposed. Various numerical approaches aremore » described to modify FLOW to include the jump condition. It is proved that the new methods converge to the correct solution even in extreme cases of very large {beta}, while they agree with the results obtained with the old implementation of FLOW in lower-{beta} equilibria.« less

Monetary effects on fear conditioning.

PubMed

Qu, Chen; Zhang, Aiyi; Chen, Qishan

2013-04-01

Previous research has found that the loss of money as a negative secondary reinforcer was as effective as a primary reinforcer during fear conditioning. The purpose of the present study was to explore the effect of monetary gain as a positive secondary reinforcer in fear conditioning. Participants were assigned to a high-reward group or low-reward group. Three kinds of squares prompting non-compensation shock, compensation shock, and no shock were presented. Skin conductance responses (SCRs) and self-ratings were recorded. The results revealed that (a) both SCRs and self-ratings in the compensation shock condition were lower than in the non-compensation shock condition, suggesting that money might block the learning stage of fear conditioning; and (b) a higher ratio of fear reduction was present in self-rating when compared to SCRs, suggesting that people might overstate the utility of money, subjectively. Monetary effects, the effects of different amounts of money, and the differences between subjective and physiological levels are discussed.

Effects of sucrose concentration and water deprivation on Pavlovian conditioning and responding for conditioned reinforcement.

PubMed

Tabbara, Rayane I; Maddux, Jean-Marie N; Beharry, Priscilla F; Iannuzzi, Jessica; Chaudhri, Nadia

2016-04-01

An appetitive Pavlovian conditioned stimulus (CS) can predict an unconditioned stimulus (US) and acquire incentive salience. We tested the hypothesis that US intensity and motivational state of the subject would influence Pavlovian learning and impact the attribution of incentive salience to an appetitive Pavlovian CS. To this end, we examined the effects of sucrose concentration and water deprivation on the acquisition of Pavlovian conditioning and responding for a conditioned reinforcer. Male Long-Evans rats (Harlan; 220-240 g) receiving 3% (3S) or 20% (20S) sucrose were either non-water deprived or given water for 1 hr per day. During Pavlovian conditioning sessions, half the rats in each concentration and deprivation condition received a 10-s CS paired with 0.2 ml of sucrose (16 trials/session; 3.2 ml/session). The remainder received unpaired CS and US presentations. Entries into a port where sucrose was delivered were recorded. Next, responding for conditioned reinforcement was tested, wherein pressing an active lever produced the CS and pressing an inactive lever had no consequences. CS-elicited port entries increased, and latency to the first CS-elicited port entry decreased across sessions in paired groups. Water deprivation augmented these effects, whereas sucrose concentration had no significant impact on behavior. Responding for conditioned reinforcement was observed in the 20S water-deprived, paired group. Thus, water deprivation can facilitate the acquisition of Pavlovian conditioning, potentially by enhancing motivational state, and a high-intensity US and a high motivational state can interact to heighten the attribution of incentive salience to an appetitive Pavlovian CS. (PsycINFO Database Record (c) 2016 APA, all rights reserved).

Conditioning biomass for microbial growth

DOEpatents

Bodie, Elizabeth A; England, George

2015-03-31

The present invention relates to methods for improving the yield of microbial processes that use lignocellulose biomass as a nutrient source. The methods comprise conditioning a composition comprising lignocellulose biomass with an enzyme composition that comprises a phenol oxidizing enzyme. The conditioned composition can support a higher rate of growth of microorganisms in a process. In one embodiment, a laccase composition is used to condition lignocellulose biomass derived from non-woody plants, such as corn and sugar cane. The invention also encompasses methods for culturing microorganisms that are sensitive to inhibitory compounds in lignocellulose biomass. The invention further provides methods of making a product by culturing the production microorganisms in conditioned lignocellulose biomass.

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

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

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

Baek, Joo S.; Diamond, David

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

Shared decision making among parents of children with mental health conditions compared to children with chronic physical conditions.

PubMed

Butler, Ashley M; Elkins, Sara; Kowalkowski, Marc; Raphael, Jean L

2015-02-01

High quality care in pediatrics involves shared decision making (SDM) between families and providers. The extent to which children with common mental health disorders experience SDM is not well known. The objectives of this study were to examine how parent-reported SDM varies by child health (physical illness, mental health condition, and comorbid mental and physical conditions) and to examine whether medical home care attenuates any differences. We analyzed data on children (2-17 years) collected through the 2009/2010 National Survey of Children with Special Health Care Needs. The sample consisted of parents of children in one of three child health categories: (1) children with a chronic physical illness but no mental health condition; (2) children with a common mental health condition but no chronic physical condition; and (3) children with comorbid mental and chronic physical conditions. The primary dependent variable was parent-report of provider SDM. The primary independent variable was health condition category. Multivariate linear regression analyses were conducted. Multivariate analyses controlling for sociodemographic variables and parent-reported health condition impact indicated lower SDM among children with a common mental health condition-only (B = -0.40; p < 0.01) and children with comorbid conditions (B = -0.67; p < 0.01) compared to children with a physical condition-only. Differences in SDM for children with a common mental health condition-only were no longer significant in the model adjusting for medical home care. However, differences in SDM for children with comorbid conditions persisted after adjusting for medical home care. Increasing medical home care may help mitigate differences in SDM for children with mental health conditions-only. Other interventions may be needed to improve SDM among children with comorbid mental and physical conditions.

Posterior consistency in conditional distribution estimation

PubMed Central

Pati, Debdeep; Dunson, David B.; Tokdar, Surya T.

2014-01-01

A wide variety of priors have been proposed for nonparametric Bayesian estimation of conditional distributions, and there is a clear need for theorems providing conditions on the prior for large support, as well as posterior consistency. Estimation of an uncountable collection of conditional distributions across different regions of the predictor space is a challenging problem, which differs in some important ways from density and mean regression estimation problems. Defining various topologies on the space of conditional distributions, we provide sufficient conditions for posterior consistency focusing on a broad class of priors formulated as predictor-dependent mixtures of Gaussian kernels. This theory is illustrated by showing that the conditions are satisfied for a class of generalized stick-breaking process mixtures in which the stick-breaking lengths are monotone, differentiable functions of a continuous stochastic process. We also provide a set of sufficient conditions for the case where stick-breaking lengths are predictor independent, such as those arising from a fixed Dirichlet process prior. PMID:25067858

Eyeblink conditioning in the developing rabbit

PubMed Central

Brown, Kevin L.; Woodruff-Pak, Diana S.

2011-01-01

Eyeblink classical conditioning in pre-weanling rabbits was examined in the present study. Using a custom lightweight headpiece and restrainer, New Zealand white littermates were trained once daily in 400 ms delay eyeblink classical conditioning from postnatal days (PD) 17–21 or PD 24–28. These ages were chosen because eyeblink conditioning emerges gradually over PD 17–24 in rats (Stanton, Freeman, & Skelton, 1992), another altricial species with neurodevelopmental features similar to those of rabbits. Consistent with well-established findings in rats, rabbits trained from PD 24–28 showed greater conditioning relative to littermates trained from PD 17–21. Both age groups displayed poor retention of eyeblink conditioning at retraining one month after acquisition. These findings are the first to demonstrate eyeblink conditioning in the developing rabbit. With further characterization of optimal conditioning parameters, this preparation may have applications to neurodevelopmental disease models as well as research exploring the ontogeny of memory. PMID:21953433

Body temperature as a conditional response measure for pavlovian fear conditioning.

PubMed

Godsil, B P; Quinn, J J; Fanselow, M S

2000-01-01

On six days rats were exposed to each of two contexts. They received an electric shock in one context and nothing in the other. Rats were tested later in each environment without shock. The rats froze and defecated more often in the shock-paired environment; they also exhibited a significantly larger elevation in rectal temperature in that environment. The rats discriminated between each context, and we suggest that the elevation in temperature is the consequence of associative learning. Thus, body temperature can be used as a conditional response measure in Pavlovian fear conditioning experiments that use footshock as the unconditional stimulus.

Plant Condition Remote Monitoring Technique

NASA Technical Reports Server (NTRS)

Fotedar, L. K.; Krishen, K.

1996-01-01

This paper summarizes the results of a radiation transfer study conducted on houseplants using controlled environmental conditions. These conditions included: (1) air and soil temperature; (2) incident and reflected radiation; and (3) soil moisture. The reflectance, transmittance, and emittance measurements were conducted in six spectral bands: microwave, red, yellow, green, violet and infrared, over a period of three years. Measurements were taken on both healthy and diseased plants. The data was collected on plants under various conditions which included: variation in plant bio-mass, diurnal variation, changes in plant pathological conditions (including changes in water content), different plant types, various disease types, and incident light wavelength or color. Analysis of this data was performed to yield an algorithm for plant disease from the remotely sensed data.

Air regenerating and conditioning

NASA Technical Reports Server (NTRS)

Grishayenkov, B. G.

1975-01-01

Various physicochemical methods of regenerating and conditioning air for spacecraft are described with emphasis on conditions which affect efficiency of the system. Life support systems used in closed, hermetically sealed environments are discussed with references to actual application in the Soviet Soyuz and Voskhod manned spacecraft. Temperature and humidity control, removal of carbon dioxide, oxygen regeneration, and removal of bacteria and viruses are among the factors considered.

Conditioned stimulus control in the rat circadian system depends on clock resetting during conditioning.

PubMed

Arvanitogiannis, A; Amir, S

1999-12-01

The authors examined the ability of a conditioned stimulus (CS; mild air disturbance) previously paired with an entraining light pulse to reset the circadian pacemaker in rats. Rats were entrained to a single 30-min light stimulus delivered every 25 hr or 24 hr (T cycle). Each daily light presentation was paired with the CS. After at least 20 days of stable entrainment to each of the T cycles, the rats were allowed to free run and were then presented with the CS at circadian time 15. CS-induced phase shifts in wheel-running activity rhythms were taken as evidence for conditioning. For the most part, conditioning occurred after CS-light pairings on the 25-hr but not 24-hr T cycle. The results suggest that CS control of the circadian clock phase depends on the effect that the entraining light pulse has on the clock during conditioning.

Thermal ageing and short-range ordering of Alloy 690 between 350 and 550 °C

NASA Astrophysics Data System (ADS)

Mouginot, Roman; Sarikka, Teemu; Heikkilä, Mikko; Ivanchenko, Mykola; Ehrnstén, Ulla; Kim, Young Suk; Kim, Sung Soo; Hänninen, Hannu

2017-03-01

Thermal ageing of Alloy 690 triggers an intergranular (IG) carbide precipitation and is known to promote an ordering reaction causing lattice contraction. It may affect the long-term primary water stress corrosion cracking (PWSCC) resistance of pressurized water reactor (PWR) components. Four conditions of Alloy 690 (solution annealed, cold-rolled and/or heat-treated) were aged between 350 and 550 °C for 10 000 h and characterized. Although no direct observation of ordering was made, variations in hardness and lattice parameter were attributed to the formation of short-range ordering (SRO) in all conditions with a peak level at 420 °C, consistent with the literature. Prior heat treatment induced ordering before thermal ageing. At higher temperatures, stress relaxation, recrystallization and α-Cr precipitation were observed in the cold-worked samples, while a disordering reaction was inferred in all samples based on a decrease in hardness. IG precipitation of M23C6 carbides increased with increasing ageing temperature in all conditions, as well as diffusion-induced grain boundary migration (DIGM).

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

Beam conditioning for FELs: Consequences and methods

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

Wolski, A.; Penn, G.; Sessler, A.

2004-06-29

The consequences of beam conditioning in four example cases (VISA, a Soft X-Ray FEL, LCLS and a ''Greenfield'' FEL) are examined. It is shown that in emittance limited cases, proper conditioning reduces sensitivity to the transverse emittance and, furthermore, allows for stronger focusing in the undulator. Simulations show higher saturation power, with gain lengths reduced by a factor of two or more. The beam dynamics in a general conditioning system are studied, with ''matching conditions'' derived for achieving conditioning without growth in the effective emittance. Various conditioning lattices are considered, and expressions derived for the amount of conditioning provided inmore » each case when the matching conditions are satisfied. These results show that there is no fundamental obstacle to producing beam conditioning, and that the problem can be reduced to one of proper lattice design. Nevertheless, beam conditioning will not be easy to implement in practice.« less

Do Conditional Reinforcers Count?

PubMed Central

Davison, Michael; Baum, William M

2006-01-01

Six pigeons were trained on a procedure in which seven components arranged different food-delivery ratios on concurrent variable-interval schedules each session. The components were unsignaled, lasted for 10 food deliveries, and occurred in random order with a 60-s blackout between components. The schedules were arranged using a switching-key procedure in which two responses on a center key changed the schedules and associated stimuli on two side keys. In Experiment 1, over five conditions, an increasing proportion of food deliveries accompanied by a magazine light was replaced with the presentation of the magazine light only. Local analyses of preference showed preference pulses toward the alternative that had just produced either a food-plus-magazine-light or magazine-light-only presentation, but pulses after food deliveries were always greater than those after magazine lights. Increasing proportions of magazine lights did not change the size of preference pulses after food or magazine-light presentations. Experiment 2 investigated the effects of correlations between food ratios and magazine-light ratios: In Condition 6, magazine-light ratios in components were inversely correlated (−1.0) with food ratios, and in Condition 7, magazine-light ratios were uncorrelated with food ratios. In Conditions 8 and 9, pecks also produced occasional 2.5-s flashes of a green keylight. In Condition 8, food and magazine-light ratios were correlated 1.0 whereas food and green-key ratios were correlated −1.0. In Condition 9, food and green-key ratios were correlated 1.0 whereas food and magazine-light ratios were correlated −1.0. Preference pulses toward alternatives after magazine lights and green keys depended on the correlation between these event ratios and the food ratios: If the ratios were correlated +1.0, positive preference pulses resulted; if the correlation was −1.0, preference pulses were negative. These results suggest that the Law of Effect has more to do with

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

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.

Fallon Geothermal Exploration Project, Naval Air Station, Fallon, Nevada.

DTIC Science & Technology

1980-05-01

magneto- telluric studies. LINEAMENT ANALYSIS As part of the initial phase of the Fallon Exploration Project, a composite lineament analysis of the region...Nevada. United States Geological Survey Bulletin 750, 1924, pp. 79-86. Hoover, D. B., R. M. Senterfit, and Bruce Radtke. Telluric Profile Loca- tion...Map and Telluric Data for the Salt Wells Known Geothermal Resource Area, Nevada. United States Geological Survey Open File Report 77-66F, 1977. Horton

Operation TEAPOT. Report of the Test Manager Joint Test Organization

DTIC Science & Technology

1981-11-01

Analysis of air, water, and milk samples were made at the laboratory at Mercury. z. 6.4 MONITORING PROCEDURES A mobile surface monitoring group consisting o...additional weather observations proved very use- ful in weather analysis and forecasting as well as for monitoring winds aloft prior to shot time. The...data were also valuable in post analysis for accurate plotting of fallout and determining cloud trajectories. The loca- tion and type of operations of

International Workshop on Millimeter Waves (1992) Held in Orvieto, Italy on April 22-24, 1992

DTIC Science & Technology

1992-04-24

1fillinmler4Vtlatri Circuits T. Yotwymatiat. Receott IDeehotnns. #of NRI)-;,.,s. Te, impIig It. 11. Jamset. Adv’anced Design Tee hnu jues fir Leiear... designed for the sky radiation measurement. - consideration of typical flight altitudes of 300m to Its output delivers a mean-value of the relevant...Electrolytic Processes: Anodic, Etching and Cathodic -increase of Surface to Volume Ratio metal deponition. Loca-Stuctre epoitin b pont- ikea ) no damage

Archeological Investigations in Cochiti Reservoir, New Mexico. Volume 2. Excavation and Analysis 1975 Season.

DTIC Science & Technology

1977-01-01

groups who used this ratio should be considered. them to process wild seeds and other vegeta material. The trough metate has been considered a specialized...transition and as an indi- was composed entirely of jar formswith wide oriface. cation of the importance assumed by wild vegetal ma- not as difficult to...For example. Ovis/Capra meat is con- faunal resource utilization from the excavated site loca- sidered most edible when the animal is young. Meat

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.

Neutronics Studies of Uranium-bearing Fully Ceramic Micro-encapsulated Fuel for PWRs

DOE PAGES

George, Nathan M.; Maldonado, G. Ivan; Terrani, Kurt A.; ...

2014-12-01

Our study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resultingmore » operating cycle length. Moreover, to match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO 2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO 2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.« less

Conditional Tests for Localizing Trait Genes

PubMed Central

Di, Yanming; Thompson, Elizabeth A.

2009-01-01

Background/Aims With pedigree data, genetic linkage can be detected using inheritance vector tests, which explore the discrepancy between the posterior distribution of the inheritance vectors given observed trait values and the prior distribution of the inheritance vectors. In this paper, we propose conditional inheritance vector tests for linkage localization. These conditional tests can also be used to detect additional linkage signals in the presence of previously detected causal genes. Methods For linkage localization, we propose to perform inheritance vector tests conditioning on the inheritance vectors at two positions bounding a test region. We can detect additional linkage signals by conducting a further conditional test in a region with no previously detected genes. We use randomized p values to extend the marginal and conditional tests when the inheritance vectors cannot be completely determined from genetic marker data. Results We conduct simulation studies to compare and contrast the marginal and the conditional tests and to demonstrate that randomized p values can capture both the significance and the uncertainty in the test results. Conclusions The simulation results demonstrate that the proposed conditional tests provide useful localization information, and with informative marker data, the uncertainty in randomized marginal and conditional test results is small. PMID:19439976

Aquatic conditions

Treesearch

Warren E. Heilman

1999-01-01

This publication provides citizens, private and public organizations, scientists, and others with information about the aquatic conditions in or near national forests in the Ozark-Ouachita Highlands: the Mark Twain in Missouri, the Ouachita in Arkansas and Oklahoma, and the Ozark-St. Francis National Forests in Arkansas. This report includes water quality analyses...

Quantum "violation" of Dirichlet boundary condition

NASA Astrophysics Data System (ADS)

Park, I. Y.

2017-02-01

Dirichlet boundary conditions have been widely used in general relativity. They seem at odds with the holographic property of gravity simply because a boundary configuration can be varying and dynamic instead of dying out as required by the conditions. In this work we report what should be a tension between the Dirichlet boundary conditions and quantum gravitational effects, and show that a quantum-corrected black hole solution of the 1PI action no longer obeys, in the naive manner one may expect, the Dirichlet boundary conditions imposed at the classical level. We attribute the 'violation' of the Dirichlet boundary conditions to a certain mechanism of the information storage on the boundary.

Ondansetron blocks LiCl-induced conditioned place avoidance but not conditioned taste/flavor avoidance in rats

PubMed Central

Rinaman, Linda; Saboury, Mitra; Litvina, Elizabeth

2009-01-01

The ability of an experimental agent to support conditioned taste/flavor avoidance (CT/FA) in rats often is interpreted as sufficient evidence that the agent produced a state of malaise or nausea. Paradoxically, however, CT/FA also is induced by certain drugs that support conditioned preferences in rats, suggesting that CT/FA is insufficient to reveal a negative hedonic state. The present study tested the hypothesis that the anti-nausea drug ondansetron (OND) would block the ability of nauseogenic lithium chloride (LiCl) to support conditioned place avoidance (CPA), without attenuating LiCl-induced CT/FA. After pre-treatment with either OND or vehicle, rats were conditioned with i.p. injection of 0.15M LiCl containing 2% saccharin (LiCl+sac) on conditioning day 1, and with 0.15M NaCl alone on conditioning day 2. Rats were confined to a distinct chamber of a CPA apparatus after each conditioning injection. In other rats, OND or vehicle pre-treatment was followed by NaCl+sac on conditioning day 1, and LiCl alone on day 2. Subsequent testing revealed that OND blocked the ability of LiCl to support CPA. Conversely, in the same rats, OND did not alter the ability of LiCl to condition avoidance of 0.2% sac solution during a 60 min bottle test. In a separate experiment, a sensitive 2-bottle choice test was used to confirm that OND pretreatment does not reduce the ability of LiCl to support CT/FA. These results support the view that CPA is an additional useful tool to reveal the experience of malaise and nausea in rats, whereas CT/FA demonstrated in bottle intake tests is insufficient for this purpose. PMID:19583975

Trace conditioning in insects—keep the trace!

PubMed Central

Dylla, Kristina V.; Galili, Dana S.; Szyszka, Paul; Lüdke, Alja

2013-01-01

Trace conditioning is a form of associative learning that can be induced by presenting a conditioned stimulus (CS) and an unconditioned stimulus (US) following each other, but separated by a temporal gap. This gap distinguishes trace conditioning from classical delay conditioning, where the CS and US overlap. To bridge the temporal gap between both stimuli and to form an association between CS and US in trace conditioning, the brain must keep a neural representation of the CS after its termination—a stimulus trace. Behavioral and physiological studies on trace and delay conditioning revealed similarities between the two forms of learning, like similar memory decay and similar odor identity perception in invertebrates. On the other hand differences were reported also, like the requirement of distinct brain structures in vertebrates or disparities in molecular mechanisms in both vertebrates and invertebrates. For example, in commonly used vertebrate conditioning paradigms the hippocampus is necessary for trace but not for delay conditioning, and Drosophila delay conditioning requires the Rutabaga adenylyl cyclase (Rut-AC), which is dispensable in trace conditioning. It is still unknown how the brain encodes CS traces and how they are associated with a US in trace conditioning. Insects serve as powerful models to address the mechanisms underlying trace conditioning, due to their simple brain anatomy, behavioral accessibility and established methods of genetic interference. In this review we summarize the recent progress in insect trace conditioning on the behavioral and physiological level and emphasize similarities and differences compared to delay conditioning. Moreover, we examine proposed molecular and computational models and reassess different experimental approaches used for trace conditioning. PMID:23986710

Boundary Condition for Modeling Semiconductor Nanostructures

NASA Technical Reports Server (NTRS)

Lee, Seungwon; Oyafuso, Fabiano; von Allmen, Paul; Klimeck, Gerhard

2006-01-01

A recently proposed boundary condition for atomistic computational modeling of semiconductor nanostructures (particularly, quantum dots) is an improved alternative to two prior such boundary conditions. As explained, this boundary condition helps to reduce the amount of computation while maintaining accuracy.

14 CFR 25.349 - Rolling conditions.

Code of Federal Regulations, 2013 CFR

2013-01-01

...(b): (1) Conditions corresponding to steady rolling velocities must be investigated. In addition, conditions corresponding to maximum angular acceleration must be investigated for airplanes with engines or other weight concentrations outboard of the fuselage. For the angular acceleration conditions, zero...

14 CFR 25.349 - Rolling conditions.

Code of Federal Regulations, 2014 CFR

2014-01-01

...(b): (1) Conditions corresponding to steady rolling velocities must be investigated. In addition, conditions corresponding to maximum angular acceleration must be investigated for airplanes with engines or other weight concentrations outboard of the fuselage. For the angular acceleration conditions, zero...

14 CFR 25.349 - Rolling conditions.

Code of Federal Regulations, 2012 CFR

2012-01-01

...(b): (1) Conditions corresponding to steady rolling velocities must be investigated. In addition, conditions corresponding to maximum angular acceleration must be investigated for airplanes with engines or other weight concentrations outboard of the fuselage. For the angular acceleration conditions, zero...

14 CFR 25.349 - Rolling conditions.

Code of Federal Regulations, 2011 CFR

2011-01-01

...(b): (1) Conditions corresponding to steady rolling velocities must be investigated. In addition, conditions corresponding to maximum angular acceleration must be investigated for airplanes with engines or other weight concentrations outboard of the fuselage. For the angular acceleration conditions, zero...

National Coastal Condition Report I Factsheet

EPA Pesticide Factsheets

The National Coastal Condition Report describes the ecological and environmental conditions in U.S. coastal waters. This first-of-its-kind Report, presents a broad baseline picture of the overall condition of U.S. coastal waters as fair to poor.

Extinction of conditioned opiate withdrawal in rats in a two-chambered place conditioning apparatus

PubMed Central

Myers, Karyn M.; Bechtholt-Gompf, Anita J.; Coleman, Brian R.; Carlezon, William A.

2016-01-01

Conditioned opiate withdrawal contributes to relapse in addicts and can be studied in rats using the opiate withdrawal-induced conditioned place aversion (OW-CPA) paradigm. Attenuation of conditioned withdrawal through extinction may be beneficial in the treatment of addiction. Here we describe a protocol for studying OW-CPA extinction using a two-chambered place conditioning apparatus. Rats are made dependent on morphine through subcutaneous implantation of morphine pellets and then trained to acquire OW-CPA through pairings of one chamber with naloxone-precipitated withdrawal and the other chamber with saline. Extinction training consists of re-exposures to both chambers in the absence of precipitated withdrawal. Rats tested following the completion of training show a decline in avoidance of the formerly naloxone-paired chamber with increasing numbers of extinction training sessions. The protocol takes a minimum of seven days; the exact duration varies with the amount of extinction training, which is determined by the goals of the experiment. PMID:22362157

Benign Breast Problems and Conditions

MedlinePlus

... a benign breast condition? • What is breast self-awareness? • Glossary What is a benign breast condition? A ... risks, and test results. What is breast self-awareness? Being aware of how your breasts normally look ...

Electron-beam conditioning by thomson scattering.

PubMed

Schroeder, C B; Esarey, E; Leemans, W P

2004-11-05

A method is proposed for conditioning electron beams via Thomson scattering. The conditioning provides a quadratic correlation between the electron energy deviation and the betatron amplitude of the electrons, which results in enhanced gain in free-electron lasers. Quantum effects imply conditioning must occur at high laser fluence and moderate electron energy. Conditioning of x-ray free-electron lasers should be achievable with present laser technology, leading to significant size and cost reductions of these large-scale facilities.

Lightning-channel conditioning

NASA Astrophysics Data System (ADS)

Sonnenfeld, R.; da Silva, C. L.; Eack, K.; Edens, H. E.; Harley, J.; McHarg, M.; Contreras Vidal, L.

2017-12-01

The concept of "conditioning" has several distinct applications in understanding lightning. It is commonly associated to the greater speed of dart-leaders vs. stepped leaders and the retrace of a cloud-to-ground channel by later return strokes. We will showadditional examples of conditioning: (A) High-speed videos of triggered flashes show "dark" periods of up to 50 ms between rebrightenings of an existing channel. (B) Interferometer (INTF) images of intra-cloud (IC) flashes demonstrate that electric-field "K-changes" correspond to rapid propagation of RF impulses along a previously formed channel separated by up to 20 ms with little RF emission on that channel. (C) Further, INTF images (like the one below) frequently show that the initial IC channel is more branched and "fuzzier'' than its later incarnations. Also, we contrast high-speed video, INTF observations, and spectroscopic measurements with possible physical mechanisms that can explain how channel conditioning guides and facilitates dart leader propagation. These mechanisms include: (1) a plasmochemical effect where electrons are stored in negative ions and released during the dart leader propagation via field-induced detachment; (2) small-amplitude residual currents that can maintain electrical conductivity; and (3) slow heat conduction cooling of plasma owing to channel expansion dynamics.

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

The Near Earth Asteroid Medical Conditions List

NASA Technical Reports Server (NTRS)

Barr, Yael R.; Watkins, S. D.

2011-01-01

Purpose: The Exploration Medical Capability (ExMC) element is one of six elements within NASA s Human Research Program (HRP) and is responsible for addressing the risk of "the inability to adequately recognize or treat an ill or injured crewmember" for exploration-class missions. The Near Earth Asteroid (NEA) Medical Conditions List, constructed by ExMC, is the first step in addressing the above-mentioned risk for the 13-month long NEA mission. The NEA mission is being designed by NASA's Human Space Flight Architecture Team (HAT). The purpose of the conditions list is to serve as an evidence-based foundation for determining which medical conditions could affect a crewmember during the NEA mission, which of those conditions would be of concern and require treatment, and for which conditions a gap in knowledge or technology development exists. This information is used to focus research efforts and technology development to ensure that the appropriate medical capabilities are available for exploration-class missions. Scope and Approach: The NEA Medical Conditions List is part of a broader Space Medicine Exploration Medical Conditions List (SMEMCL), which incorporates various exploration-class design reference missions (DRMs). The conditions list contains 85 medical conditions which could occur during space flight and which are derived from several sources: Long-Term Surveillance of Astronaut Health (LSAH) in-flight occurrence data, The Space Shuttle (STS) Medical Checklist, The International Space Station (ISS) Medical Checklist, and subject matter expert opinion. Each medical condition listed has been assigned a clinical priority and a clinical priority rationale based on incidence, consequence, and mitigation capability. Implementation: The conditions list is a "living document" and as such, new conditions can be added to the list, and the priority of conditions on the list can be adjusted as the DRM changes, and as screening, diagnosis, or treatment capabilities

Contextual fear conditioning in zebrafish.

PubMed

Kenney, Justin W; Scott, Ian C; Josselyn, Sheena A; Frankland, Paul W

2017-10-01

Zebrafish are a genetically tractable vertebrate that hold considerable promise for elucidating the molecular basis of behavior. Although numerous recent advances have been made in the ability to precisely manipulate the zebrafish genome, much less is known about many aspects of learning and memory in adult fish. Here, we describe the development of a contextual fear conditioning paradigm using an electric shock as the aversive stimulus. We find that contextual fear conditioning is modulated by shock intensity, prevented by an established amnestic agent (MK-801), lasts at least 14 d, and exhibits extinction. Furthermore, fish of various background strains (AB, Tu, and TL) are able to acquire fear conditioning, but differ in fear extinction rates. Taken together, we find that contextual fear conditioning in zebrafish shares many similarities with the widely used contextual fear conditioning paradigm in rodents. Combined with the amenability of genetic manipulation in zebrafish, we anticipate that our paradigm will prove to be a useful complementary system in which to examine the molecular basis of vertebrate learning and memory. © 2017 Kenney et al.; Published by Cold Spring Harbor Laboratory Press.

Pavlovian conditioning of nausea and vomiting.

PubMed

Stockhorst, Ursula; Steingrueber, Hans-Joachim; Enck, Paul; Klosterhalfen, Sibylle

2006-10-30

Cancer patients undergoing cytotoxic drug treatment often experience side-effects, the most distressing being nausea and vomiting. Despite antiemetic drugs, 25-30% of the chemotherapy patients report these side-effects when being re-exposed to the stimuli that usually signal the chemotherapy session and its drug infusion. These symptoms are called anticipatory nausea and anticipatory vomiting. The present paper summarizes the evidence that anticipatory vomiting is acquired by Pavlovian conditioning, and, consequently, may be alleviated by conditioning techniques. To explore the mechanisms that induce and alleviate conditioned nausea and vomiting further, a conditioned nausea model was established in healthy humans using body rotation as the nausea-inducing treatment. The validity of this motion sickness model to examine conditioning mechanisms in the acquisition and alleviation of conditioned nausea was demonstrated. Cortisol and tumor-necrosis factor-alpha were elevated as endocrine and immunological correlates of nausea. Data in the rotation-induced motion sickness model indicated that gender is an important moderator variable to be considered in further studies. The paper concludes with a review of applications of the demonstrated conditioning principles as interventions to ameliorate distressing anticipatory nausea or anticipatory vomiting in cancer patients undergoing chemotherapy.

National Wetland Condition Assessment 2011: A ...

EPA Pesticide Factsheets

The National Wetland Condition Assessment 2011: A Collaborative Survey presents the results of an unprecedented assessment of the nation’s wetlands. This report is part of the National Aquatic Resource Surveys, a series of statistically based surveys designed to provide the public and decision makers with nationally consistent and representative information on the condition of all the nation's waters. The National Wetland Condition report provides information on the biological condition of the nation’s wetlands and key stressors that affect them.

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.

9 CFR 102.6 - Conditional licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... PRODUCTS § 102.6 Conditional licenses. In order to meet an emergency condition, limited market, local...) of this part, issue a conditional U.S. Veterinary Biological Product License to an establishment... available, the Administrator shall either reissue the U.S. Veterinary Biological Product License or allow it...

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

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.

The National Wetland Condition Assessment

EPA Science Inventory

The first National Wetland Condition Assessment (NWCA) was conducted in 2011 by the US Environmental Protection Agency (USEPA). Vegetation, algae, soil, water chemistry,and hydrologic data were collected at each of 1138 sites across the contiguous US. Ecological condition was ass...

16 CFR 1203.8 - Conditioning environments.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 16 Commercial Practices 2 2012-01-01 2012-01-01 false Conditioning environments. 1203.8 Section... SAFETY STANDARD FOR BICYCLE HELMETS The Standard § 1203.8 Conditioning environments. Helmets shall be conditioned to one of the following environments prior to testing in accordance with the test schedule at...

16 CFR 1203.8 - Conditioning environments.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 16 Commercial Practices 2 2011-01-01 2011-01-01 false Conditioning environments. 1203.8 Section... SAFETY STANDARD FOR BICYCLE HELMETS The Standard § 1203.8 Conditioning environments. Helmets shall be conditioned to one of the following environments prior to testing in accordance with the test schedule at...

16 CFR 1203.8 - Conditioning environments.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 16 Commercial Practices 2 2014-01-01 2014-01-01 false Conditioning environments. 1203.8 Section... SAFETY STANDARD FOR BICYCLE HELMETS The Standard § 1203.8 Conditioning environments. Helmets shall be conditioned to one of the following environments prior to testing in accordance with the test schedule at...

16 CFR 1203.8 - Conditioning environments.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Conditioning environments. 1203.8 Section... SAFETY STANDARD FOR BICYCLE HELMETS The Standard § 1203.8 Conditioning environments. Helmets shall be conditioned to one of the following environments prior to testing in accordance with the test schedule at...

OPAL Land Condition Model

DTIC Science & Technology

2014-08-01

ER D C/ CE RL S R- 14 -7 Optimal Allocation of Land for Training and Non-training Uses OPAL Land Condition Model Co ns tr uc tio n En...Optimal Allocation of Land for Training and Non-training Uses ERDC/CERL SR-14-7 August 2014 OPAL Land Condition Model Daniel Koch, Scott Tweddale...programmer information supporting the Op- timal Programming of Army Lands ( OPAL ) model, which was designed for use by trainers, Integrated Training

10 CFR 950.10 - Conditional agreement.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Conditional agreement. 950.10 Section 950.10 Energy....10 Conditional agreement. (a) Purpose. The Department and a sponsor may enter into a Conditional Agreement. The Department will enter into a Standby Support Contract with the first six sponsors to satisfy...

10 CFR 950.10 - Conditional agreement.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Conditional agreement. 950.10 Section 950.10 Energy....10 Conditional agreement. (a) Purpose. The Department and a sponsor may enter into a Conditional Agreement. The Department will enter into a Standby Support Contract with the first six sponsors to satisfy...

40 CFR 91.311 - Test conditions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... engine air at the inlet to the engine and the dry atmospheric pressure (designated as p s and expressed... rates at standard conditions for temperature and pressure. Use these conditions consistently throughout all calculations. Standard conditions for temperature and pressure are 25 °C and 101.3 kPa. (b) Engine...

29 CFR 1926.1402 - Ground conditions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 29 Labor 8 2011-07-01 2011-07-01 false Ground conditions. 1926.1402 Section 1926.1402 Labor... Ground conditions. (a) Definitions. (1) “Ground conditions” means the ability of the ground to support...) The equipment must not be assembled or used unless ground conditions are firm, drained, and graded to...

Nutritional condition of Northern Yellowstone Elk

USGS Publications Warehouse

Cook, R.C.; Cook, J.G.; Mech, L.D.

2004-01-01

Ultrasonography and body condition scoring was used to estimate nutritional condition of northern Yellowstone elk in late winter. Probability of pregnancy was related to body fat, and lactating cows had 50% less fat than non-lactating cows. For mild to normal winters, most of the elk were in good condition.

Effects of early developmental conditions on innate immunity are only evident under favourable adult conditions in zebra finches

NASA Astrophysics Data System (ADS)

de Coster, Greet; Verhulst, Simon; Koetsier, Egbert; de Neve, Liesbeth; Briga, Michael; Lens, Luc

2011-12-01

Long-term effects of unfavourable conditions during development can be expected to depend on the quality of the environment experienced by the same individuals during adulthood. Yet, in the majority of studies, long-term effects of early developmental conditions have been assessed under favourable adult conditions only. The immune system might be particularly vulnerable to early environmental conditions as its development, maintenance and use are thought to be energetically costly. Here, we studied the interactive effects of favourable and unfavourable conditions during nestling and adult stages on innate immunity (lysis and agglutination scores) of captive male and female zebra finches ( Taeniopygia guttata). Nestling environmental conditions were manipulated by a brood size experiment, while a foraging cost treatment was imposed on the same individuals during adulthood. This combined treatment showed that innate immunity of adult zebra finches is affected by their early developmental conditions and varies between both sexes. Lysis scores, but not agglutination scores, were higher in individuals raised in small broods and in males. However, these effects were only present in birds that experienced low foraging costs. This study shows that the quality of the adult environment may shape the long-term consequences of early developmental conditions on innate immunity, as long-term effects of nestling environment were only evident under favourable adult conditions.

Effective surface and boundary conditions for heterogeneous surfaces with mixed boundary conditions

NASA Astrophysics Data System (ADS)

Guo, Jianwei; Veran-Tissoires, Stéphanie; Quintard, Michel

2016-01-01

To deal with multi-scale problems involving transport from a heterogeneous and rough surface characterized by a mixed boundary condition, an effective surface theory is developed, which replaces the original surface by a homogeneous and smooth surface with specific boundary conditions. A typical example corresponds to a laminar flow over a soluble salt medium which contains insoluble material. To develop the concept of effective surface, a multi-domain decomposition approach is applied. In this framework, velocity and concentration at micro-scale are estimated with an asymptotic expansion of deviation terms with respect to macro-scale velocity and concentration fields. Closure problems for the deviations are obtained and used to define the effective surface position and the related boundary conditions. The evolution of some effective properties and the impact of surface geometry, Péclet, Schmidt and Damköhler numbers are investigated. Finally, comparisons are made between the numerical results obtained with the effective models and those from direct numerical simulations with the original rough surface, for two kinds of configurations.

On High-Order Radiation Boundary Conditions

NASA Technical Reports Server (NTRS)

Hagstrom, Thomas

1995-01-01

In this paper we develop the theory of high-order radiation boundary conditions for wave propagation problems. In particular, we study the convergence of sequences of time-local approximate conditions to the exact boundary condition, and subsequently estimate the error in the solutions obtained using these approximations. We show that for finite times the Pade approximants proposed by Engquist and Majda lead to exponential convergence if the solution is smooth, but that good long-time error estimates cannot hold for spatially local conditions. Applications in fluid dynamics are also discussed.

Conditional clustering of temporal expression profiles

PubMed Central

Wang, Ling; Montano, Monty; Rarick, Matt; Sebastiani, Paola

2008-01-01

Background Many microarray experiments produce temporal profiles in different biological conditions but common cluster techniques are not able to analyze the data conditional on the biological conditions. Results This article presents a novel technique to cluster data from time course microarray experiments performed across several experimental conditions. Our algorithm uses polynomial models to describe the gene expression patterns over time, a full Bayesian approach with proper conjugate priors to make the algorithm invariant to linear transformations, and an iterative procedure to identify genes that have a common temporal expression profile across two or more experimental conditions, and genes that have a unique temporal profile in a specific condition. Conclusion We use simulated data to evaluate the effectiveness of this new algorithm in finding the correct number of clusters and in identifying genes with common and unique profiles. We also use the algorithm to characterize the response of human T cells to stimulations of antigen-receptor signaling gene expression temporal profiles measured in six different biological conditions and we identify common and unique genes. These studies suggest that the methodology proposed here is useful in identifying and distinguishing uniquely stimulated genes from commonly stimulated genes in response to variable stimuli. Software for using this clustering method is available from the project home page. PMID:18334028

10 CFR 950.10 - Conditional agreement.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Conditional agreement. 950.10 Section 950.10 Energy....10 Conditional agreement. (a) Purpose. The Department and a sponsor may enter into a Conditional... docketed by the Commission pursuant to 10 CFR part 52, and if applicable, an electronic copy of the design...

10 CFR 950.10 - Conditional agreement.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Conditional agreement. 950.10 Section 950.10 Energy....10 Conditional agreement. (a) Purpose. The Department and a sponsor may enter into a Conditional... docketed by the Commission pursuant to 10 CFR part 52, and if applicable, an electronic copy of the design...

40 CFR 90.311 - Test conditions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... pressure, and use these conditions consistently throughout all calculations. Standard conditions for temperature and pressure are 25 °C and 101.3 kPa. (b) Engine test conditions. Measure the absolute temperature (designated as T and expressed in Kelvin) of the engine air at the inlet to the engine and the dry atmospheric...

77 FR 4469 - Dental Conditions

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-30

... DEPARTMENT OF VETERANS AFFAIRS 38 CFR Part 3 RIN 2900-AN28 Dental Conditions AGENCY: Department of... rule the proposal to amend its adjudication regulations regarding service connection of dental... Veterans Benefits Administration (VBA) for service connection of dental conditions for the purpose of...

ASSESSING THE CONDITION OF SOUTH CAROLINA'S ESTUARIES: A NEW APPROACH INVOLVING INTEGRATED MEASURES OF CONDITION

EPA Science Inventory

The South Carolina Estuarine and Coastal Assessment Program (SCECAP) was initiated in 1999 to assess the condition of the state's coastal habitats using multiple measures of water quality, sediment quality, and biological condition. Sampling has subsequently been expanded to incl...

Acoustic Conditioning System Development and Conditioning Experiments on Black Seabreams in the Xiangshan Bay Sea Ranch

NASA Astrophysics Data System (ADS)

Hu, Qingsong; Rahman, Hafiz Abd ur; Jiang, Yazhou; Zhang, Shouyu; Shentu, Jikang

2018-06-01

Attracting released hatchery-reared fish to designated areas during the growth process is vital to realize the objectives of sea ranching. Based on the bottom artificial reefs and surface kelp culture facilities in the Xiangshan Bay sea ranch, we proposed systematic techniques related to acoustic conditioning of the black seabream ( Sparus macrocephalus). Experiments conducted in 12 m × 10 m × 1.6 m ponds on Xixuan Island showed that black seabream was positively sensitive to 500-600 Hz periodic signals. Conditioned responses were apparent after 8 d. Two to three days were required for recovery of the memory of a conditioned response after a 20-day interval. According to the practical application requirements in the open sea, unattended acoustic conditioning equipment was developed. The ranching equipment was used in 12 m × 12 m × 2.5 m cages, and the behavior of black seabream juveniles was successfully guided after 7 days. Of the 16000 released fish, 82.5% of them were conditioned with a flexible grading net. To avoid inducing a stress response, the juveniles were released into the sea ranch in situ from the net cage. The acoustic conditioning equipments were moved into the open sea and the aggregation phenomenon of the released fish was observed when the sound was played. After 6 months of investigation and based on Sr+ marking, only one acoustically conditioned fish was found outside the 3.5-km2 sea ranch area, thereby reached the goal of guiding activity. The practical effect in the Xiangshan Bay sea ranch showed the validity of the acoustic conditioning system, which may contribute to improve the operation of the sea ranches in the East China Sea.

Residence Conditions on Community Treatment Orders.

PubMed

Dawson, John; O'Reilly, Richard

2015-11-01

To identify the clinical reasons and legal authority for including a residential placement condition in a community treatment order (CTO). We describe the clinical reasons for imposing a residence condition and discuss how this is authorized by the laws of the Canadian provinces (using Ontario as the main example). A residence condition can facilitate numerous benefits, including: regular access to a person by a clinical team; continuing therapeutic relations; supervision of medication; provision of general medical care; and reduction in substance use, risks of victimization, and other unintended harm. A resident condition can be lawfully imposed when it clearly fits the purposes of the CTO legislation and stops short of authorizing detention in a community facility. In certain circumstances, a residence condition is clinically justified and a lawful aspect of a CTO.

Beam Conditioning for FELs: Consequences and Methods

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

Wolski, Andrzej; Penn, Gregory; Sessler, Andrew

2003-10-09

The consequences of beam conditioning in four example cases (VISA, a Soft X-Ray FEL, LCLS and a ''Greenfield'' FEL) are examined. It is shown that in emittance limited cases, proper conditioning reduces sensitivity to the transverse emittance, and allows stronger focusing in the undulator. Simulations show higher saturation power, with gain lengths reduced up to a factor of two. The beam dynamics in a general conditioning system are studied, with ''matching conditions'' derived for achieving conditioning without growth in effective emittance. Various conditioners are considered, and expressions derived for the amount of conditioning provided in each case when the matchingmore » conditions are satisfied. We discuss the prospects for conditioners based on laser and plasma systems.« less

Dynamical interpretation of conditional patterns

NASA Technical Reports Server (NTRS)

Adrian, R. J.; Moser, R. D.; Moin, P.

1988-01-01

While great progress is being made in characterizing the 3-D structure of organized turbulent motions using conditional averaging analysis, there is a lack of theoretical guidance regarding the interpretation and utilization of such information. Questions concerning the significance of the structures, their contributions to various transport properties, and their dynamics cannot be answered without recourse to appropriate dynamical governing equations. One approach which addresses some of these questions uses the conditional fields as initial conditions and calculates their evolution from the Navier-Stokes equations, yielding valuable information about stability, growth, and longevity of the mean structure. To interpret statistical aspects of the structures, a different type of theory which deals with the structures in the context of their contributions to the statistics of the flow is needed. As a first step toward this end, an effort was made to integrate the structural information from the study of organized structures with a suitable statistical theory. This is done by stochastically estimating the two-point conditional averages that appear in the equation for the one-point probability density function, and relating the structures to the conditional stresses. Salient features of the estimates are identified, and the structure of the one-point estimates in channel flow is defined.

Conditional data watchpoint management

DOEpatents

Burdick, Dean Joseph; Vaidyanathan, Basu

2010-08-24

A method, system and computer program product for managing a conditional data watchpoint in a set of instructions being traced is shown in accordance with illustrative embodiments. In one particular embodiment, the method comprises initializing a conditional data watchpoint and determining the watchpoint has been encountered. Upon that determination, examining a current instruction context associated with the encountered watchpoint prior to completion of the current instruction execution, further determining a first action responsive to a positive context examination; otherwise, determining a second action.

M3FT-16OR020202112 - Report on viability of hydrothermal corrosion resistant SiC/SiC Joint development

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

Katoh, Yutai; Koyanagi, Takaaki; Kiggans Jr, James O.

2016-06-30

Hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under PWR and BWR relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti-Si-C system, and SiC nanopowder sintering. Most of the formed joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession rate of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing activity environments. The SiC nanopowder sinteredmore » joint was the most corrosion tolerant under the oxidizing activity environment among the four joints.« less

7 CFR 29.6008 - Condition.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Condition. 29.6008 Section 29.6008 Agriculture Regulations of the Department of Agriculture AGRICULTURAL MARKETING SERVICE (Standards, Inspections, Marketing... INSPECTION Standards Definitions § 29.6008 Condition. The state of tobacco which results from the method of...

7 CFR 29.3014 - Condition.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Condition. 29.3014 Section 29.3014 Agriculture Regulations of the Department of Agriculture AGRICULTURAL MARKETING SERVICE (Standards, Inspections, Marketing... Condition. The state of tobacco which results from the method of preparation or from the degree of...

7 CFR 29.2260 - Condition.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Condition. 29.2260 Section 29.2260 Agriculture Regulations of the Department of Agriculture AGRICULTURAL MARKETING SERVICE (Standards, Inspections, Marketing... Condition. The state of tobacco which results from the method of preparation or from the degree of...

24 CFR 902.30 - Financial condition assessment.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Financial condition assessment. 902... DEVELOPMENT PUBLIC HOUSING ASSESSMENT SYSTEM PHAS Indicator #2: Financial Condition § 902.30 Financial condition assessment. (a) Objective. The objective of the Financial Condition Indicator is to measure the...

24 CFR 902.20 - Physical condition assessment.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 24 Housing and Urban Development 4 2013-04-01 2013-04-01 false Physical condition assessment. 902... DEVELOPMENT PUBLIC HOUSING ASSESSMENT SYSTEM Physical Condition Indicator § 902.20 Physical condition assessment. (a) Objective. The objective of the physical condition indicator is to determine whether a PHA is...

24 CFR 902.20 - Physical condition assessment.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 24 Housing and Urban Development 4 2012-04-01 2012-04-01 false Physical condition assessment. 902... DEVELOPMENT PUBLIC HOUSING ASSESSMENT SYSTEM Physical Condition Indicator § 902.20 Physical condition assessment. (a) Objective. The objective of the physical condition indicator is to determine whether a PHA is...

24 CFR 902.20 - Physical condition assessment.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 24 Housing and Urban Development 4 2011-04-01 2011-04-01 false Physical condition assessment. 902... DEVELOPMENT PUBLIC HOUSING ASSESSMENT SYSTEM Physical Condition Indicator § 902.20 Physical condition assessment. (a) Objective. The objective of the physical condition indicator is to determine whether a PHA is...

24 CFR 902.20 - Physical condition assessment.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 24 Housing and Urban Development 4 2014-04-01 2014-04-01 false Physical condition assessment. 902... DEVELOPMENT PUBLIC HOUSING ASSESSMENT SYSTEM Physical Condition Indicator § 902.20 Physical condition assessment. (a) Objective. The objective of the physical condition indicator is to determine whether a PHA is...

Assessing ecological conditions in the Great Lakes Connecting Channels using National Coastal Condition Assessment protocol

EPA Science Inventory

The EPA Office of Water’s National Coastal Condition Assessment (NCCA) helps satisfy the assessment and antidegradation provisions of the Clean Water Act by estimating water, sediment, and benthic quality conditions in the Great Lakes nearshore on a five-year cycle starting...

[Working conditions, living conditions and physical health problems declared among penitentiary administration personnel in France].

PubMed

Goldberg, P; Landre, M F; David, S; Goldberg, M; Dassa, S; Marne, M J

1996-06-01

A cross-sectional epidemiological survey was conducted among prison staff in France to investigate the relationships between working conditions and health. The sample included men and women 20 to 64 years old belonging to all categories of prison personnel: prison guards, administrative staff, socioeducational workers, technicians, health care workers, and managers (n = 4587, response rate 45.7%). A mailed self-administered questionnaire was used to assess sociodemographic characteristics, working conditions, and physical and mental disorders. Multiple logistic regression analyses were conducted to determine the effects of working conditions and social relationships on health of prison staff. However, the results reported here only concern 17 health disorders: body mass index, sick leave, medication use, accidents, digestive disorders, lower extremities and back disorders, hypertension, hemorrhoids, arthritis, skin disorders, urinary infections, chronic bronchitis, cholesterol, gastric ulcer, respiratory infections, ocular disorders. The living non professional conditions mostly associated with health disorders were financial difficulties (OR: 1.9 for digestive disorders, 1.8 for gastric ulcer, 1.7 for medication use) and irregularity of meals (OR = 1.5 for digestive disorders, and hypertension). In the occupational environment, the factors most associated with health disorders are seniority (OR = 4.2 for arthritis, 2.3 for cholesterol) and constraints (OR = 1.7 for lower extremities disorders). In spite of some limits associated to this kind of study, relationships between occupational and non occupational factors and physical health conditions were observed; the results also pointed out the protective role of the social relationships for health conditions.

7 CFR 989.58 - Natural condition raisins.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 8 2010-01-01 2010-01-01 false Natural condition raisins. 989.58 Section 989.58... GROWN IN CALIFORNIA Order Regulating Handling Grade and Condition Standards § 989.58 Natural condition raisins. (a) Regulation. No handler shall acquire or receive natural condition raisins which fail to meet...

7 CFR 989.58 - Natural condition raisins.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 7 Agriculture 8 2011-01-01 2011-01-01 false Natural condition raisins. 989.58 Section 989.58... GROWN IN CALIFORNIA Order Regulating Handling Grade and Condition Standards § 989.58 Natural condition raisins. (a) Regulation. No handler shall acquire or receive natural condition raisins which fail to meet...

24 CFR 902.20 - Physical condition assessment.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Physical condition assessment. 902... DEVELOPMENT PUBLIC HOUSING ASSESSMENT SYSTEM PHAS Indicator #1: Physical Condition § 902.20 Physical condition assessment. (a) Objective. The objective of the Physical Condition Indicator is to determine whether a PHA is...

Conditionally Active Min-Max Limit Regulators

NASA Technical Reports Server (NTRS)

Garg, Sanjay (Inventor); May, Ryan D. (Inventor)

2017-01-01

A conditionally active limit regulator may be used to regulate the performance of engines or other limit regulated systems. A computing system may determine whether a variable to be limited is within a predetermined range of a limit value as a first condition. The computing system may also determine whether a current rate of increase or decrease of the variable to be limited is great enough that the variable will reach the limit within a predetermined period of time with no other changes as a second condition. When both conditions are true, the computing system may activate a simulated or physical limit regulator.

45 CFR 690.124 - Conditions.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 45 Public Welfare 3 2014-10-01 2014-10-01 false Conditions. 690.124 Section 690.124 Public Welfare Regulations Relating to Public Welfare (Continued) NATIONAL SCIENCE FOUNDATION PROTECTION OF HUMAN SUBJECTS § 690.124 Conditions. With respect to any research project or any class of research projects the...