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

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...

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

Barry, Kenneth

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

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.

77 FR 37795 - Airworthiness Directives; Dassault Aviation Airplanes

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-25

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

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

PubMed

Phillips, Joanne; King, Ross; Skouteris, Helen

2014-11-01

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

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

Code of Federal Regulations, 2012 CFR

2012-07-01

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

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

Code of Federal Regulations, 2014 CFR

2014-07-01

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

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

Code of Federal Regulations, 2013 CFR

2013-07-01

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

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.

ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

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

2013-11-01

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

The effects of stainless steel radial reflector on core reactivity for small modular reactor

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

Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr; Cho, Sung Ju, E-mail: sungju@knfc.co.kr

Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5more » w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.« less

The benefits of a fast reactor closed fuel cycle in the UK

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

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

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...

ORNL rod-bundle heat-transfer test data. Volume 2. Thermal-Hydraulic Test Facility experimental data report for test 3. 03. 6AR - transient film boiling in upflow

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

Mullins, C. B.; Felde, D. K.; Sutton, A. G.

1982-04-01

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.03.6AR. This test was conducted by members of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on May 21, 1980. Objective was to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.03.6AR was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.03.6AR available. Included in the report are uncertainties in the instrument responses,more » calculated mass flows, and calculated rod powers.« less

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.

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

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

NASA Astrophysics Data System (ADS)

Alim, Fatih

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

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

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.

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.

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) on plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.« less

Prise en compte d'un couplage fin neutronique-thermique dans les calculs d'assemblage pour les reacteurs a eau pressurisee

NASA Astrophysics Data System (ADS)

Greiner, Nathan

Core simulations for Pressurized Water Reactors (PWR) is insured by a set of computer codes which allows, under certain assumptions, to approximate the physical quantities of interest, such as the effective multiplication factor or the power or temperature distributions. The neutronics calculation scheme relies on three great steps : -- the production of an isotopic cross-sections library ; -- the production of a reactor database through the lattice calculation ; -- the full-core calculation. In the lattice calculation, in which Boltzmann's transport equation is solved over an assembly geometry, the temperature distribution is uniform and constant during irradiation. This represents a set of approximations since, on the one hand, the temperature distribution in the assembly is not uniform (strong temperature gradients in the fuel pins, discrepancies between the fuel pins) and on the other hand, irradiation causes the thermal properties of the pins to change, which modifies the temperature distribution. Our work aims at implementing and introducing a neutronics-thermomechanics coupling into the lattice calculation to finely discretize the temperature distribution and to study its effects. To perform the study, CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) lattice code APOLLO2 was used for neutronics and EDF (Electricite De France) code C3THER was used for the thermal calculations. We show very small effects of the pin-scaled coupling when comparing the use of a temperature profile with the use of an uniform temperature over UOX-type and MOX-type fuels. We next investigate the thermal feedback using an assembly-scaled coupling taking into account the presence of large water gaps on an UOX-type assembly at burnup 0. We show the very small impact on the calculation of the hot spot factor. Finally, the coupling is introduced into the isotopic depletion calculation and we show that reactivity and isotopic number densities deviations remain small albeit not negligible for UOX-type and MOX-type assemblies. The specific behavior of gadolinium-stuffed fuel pins in an UO2Gd2O 3-type assembly is highlighted.

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-pixel cropland components.

Stress corrosion crack initiation of alloy 600 in PWR primary water

DOE PAGES

Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.; ...

2017-04-27

Stress corrosion crack (SCC) initiation of three mill-annealed alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and intergranular SCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Lastly, we discuss processes controlling the SCC initiation in mill-annealed alloy 600.

Stress corrosion crack initiation of alloy 600 in PWR primary water

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

Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.

Stress corrosion crack (SCC) initiation of three mill-annealed alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and intergranular SCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Lastly, we discuss processes controlling the SCC initiation in mill-annealed alloy 600.

Xenon-induced power oscillations in a generic small modular reactor

NASA Astrophysics Data System (ADS)

Kitcher, Evans Damenortey

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

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

PRESSURIZED WATER REACTOR (PWR) PROJECT TECHNICAL PROGRESS REPORT FOR THE PERIOD DECEMBER 24, 1959 TO FEBRUARY 23, 1960

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

None

< 9 A < 2 6 < 7 4 8 9 6 2 6 equalizing vent valves on air locks 2, 4, and 5 was completed. An evaluation of the failed main coolant pump No. 1-80-F-737 was completed. The design for installing combination ball check and manual stop valves on the boiler water level sight glasses, to prevent the escape of steam should a defective sight glass develop, was completed. The main coolant pumps No. 80 and No. 79 were modified by increasing the radial clearance of the impeller wear ring and by removing the upper labyrinth ring. A designmore » for relocating the cooling water flow orifice 17-J4-17 was completed. Metallurgy: Preliminary data from the Bett 69-1 in-pile thermal conductivity capsules indicate that the thermal conductivity of as-sintered ZrO/sub 2/ 34 wt.% UO/sub 2/ appears to decrease from an initial value of about 1.6 Btu/hr-ft- deg F to about 0.7 Btu/hr-ft- deg F after 17 days irradiation in an estimated perturbed flux of 4 x 10/sup 13/. The thermal conductivities of UO/sub 2/ and BeO 51 wt.% UO/sub 2/ fuel remained unchanged during this time. Examination of the two failed X-3-1 fuel plates and the two failed CR-V-m fuel plates showed that a definite burnup limitation exists for bulk UO/sub 2/i of about 16 x 10/sup 20/ to 21.5 x 10/sup 20/ fissions/cc at which point the fuel increases in volume about 4- -5%. Irradiation of both fine and coarse dis-persions of 28 wt.% UO/sub 2/in BeO to exposures of about 11 x 10/sup 20/ fissions/cc shows this material has very poor dimensional stabllity and poor fission gas retention ability. The fine particles dispersion showed approximately 4.8 times the thickness increase as did the coarse particles. Interim examination of a bulk B/sub 4/ burnable poison plate irradiated in the HB-1 loop to about 60 at.% B/sup 10/ burnup showed a 17% increase in plate thickness. The technical feasibility of fabricating blanket receptacles with full length fuel channels and an integral cover plate by form rolling was established. Hack-pressure-bonding appears to be a suitable means of incorporating void volume in fuel compartments of oxide plates. High density (99% T.D.) and improved microstructure of B/sub 4/C-SiC burnable poisons are achieved when small (2 micron) B/sub 4/C particle size powder is used ia hot pressing compacts. Measurements of the self-diffusion coefficients of uranium in UO/sub 2/ by the method of surface activity decrease were completed. Experiments on the diffusion of Xe/sup 133/ in Core 2--type UO/sup 2/ fuel platelets were completed. Diffusion anaeals carried out at 1000 deg C on samples from the X-3-1 and the 14-28 irradiation tests show that the apparent diffusion coefficient for Kr/sup 85/ incresses considerably with burnup. An average activation energy for thoron emanation in UO/sub 2/ was estimated to be 44 kcal/mole. An initial experiment on the release of helium from slightly irradiated B/sub 4/C at 900 deg C resulted in a diffusion coefficient for helium of 3.5 x 10/sup -8/ Physics: Calculatad values for seed-blanket power sharing as a function of PWR-1 Seed 1 life were compared with measured data obtained from thermal instrumentation at Shippingport. Two-dimensional depletion studies in the PWR-2 "composite cell" geometry were completed for seed assembly configurations having different radial fuel zoning. An eighth core representation is being employed for a two- dimensional depletion calculation of PWR-2. An analysis of the effect on the axial power distribution of the nonuniform temperature distribution in an 8 ft PWR-2 core loaded with 295 kg of U/sup 235/ indicated that local variations in power density of as much as 15% may occur, relative to the distribution that would exist if the axial temperature distribution were uniform. A technique was developed which makes possible an approximately correct description of the neutron capture rate within small rectangular boron wafers in diffusion theory calculations. Seed peaking factors measured in a five-cluster slab of PWR-2 mock- up materials were measured and compared with calculated peaking factors obtained using the nuclear« less

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

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

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

Assessment for advanced fuel cycle options in CANDU

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

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

2013-07-01

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

Noninvasive and Real-Time Plasmon Waveguide Resonance Thermometry

PubMed Central

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

2015-01-01

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

Structural Integrity of Water Reactor Pressure Boundary Components.

DTIC Science & Technology

1981-02-20

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

Irradiation-induced microchemical changes in highly irradiated 316 stainless steel

NASA Astrophysics Data System (ADS)

Fujii, K.; Fukuya, K.

2016-02-01

Cold-worked 316 stainless steel specimens irradiated to 74 dpa in a pressurized water reactor (PWR) were analyzed by atom probe tomography (APT) to extend knowledge of solute clusters and segregation at higher doses. The analyses confirmed that those clusters mainly enriched in Ni-Si or Ni-Si-Mn were formed at high number density. The clusters were divided into three types based on their size and Mn content; small Ni-Si clusters (3-4 nm in diameter), and large Ni-Si and Ni-Si-Mn clusters (8-10 nm in diameter). The total cluster number density was 7.7 × 1023 m-3. The fraction of large clusters was almost 1/10 of the total density. The average composition (in at%) for small clusters was: Fe, 54; Cr, 12; Mn, 1; Ni, 22; Si, 11; Mo, 1, and for large clusters it was: Fe, 44; Cr, 9; Mn, 2; Ni, 29; Si, 14; Mo,1. It was likely that some of the Ni-Si clusters correspond to γ‧ phase precipitates while the Ni-Si-Mn clusters were precursors of G phase precipitates. The APT analyses at grain boundaries confirmed enrichment of Ni, Si, P and Cu and depletion of Fe, Cr, Mo and Mn. The segregation behavior was consistent with previous knowledge of radiation induced segregation.

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

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.

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

NASA Astrophysics Data System (ADS)

Hartini, Entin; Andiwijayakusuma, Dinan

2014-09-01

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

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

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

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

2014-09-30

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

Characterization of carbon-14 generated by the nuclear power industry. Final report

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

Eabry, S.; Vance, J.N.; Cline, J.E.

1995-11-01

This report describes an evaluation of C-14 production rates in light-water reactors (LWRs) and characterization of its chemical speciation and environmental behavior. The study estimated the total production rate of the nuclide in operating PWRs and BWRs along with the assessment of the C-14 content of solid radwaste. The major source of production of C-14 in both PWR`s and BWRs was the activation of 0-17 in the water molecule and of N-14 dissolved in reactor coolant. The production of C-14 was estimated to range from 7 Ci/GW(e)-year to 11 Ci/GW(e)-year. The estimated range of the quantity of C-14 in LLWmore » was 1-2 Ci/ reactor-year which compares favorably with data obtained from shipping manifests. The environmental behavior of C-14 associated with low-level waste (LLW) disposal is greatly dependent upon its chemical speciation. This scoping study was performed to help identify the occurrence of inorganic and organic forms of C-14 in reactor coolant water and in primary coolant demineralization resins. These represent the major source for C-14 in LLW from nuclear power stations. Also, the behavior of inorganic and two of the organic forms of C-14 on soil uptake was determined by measuring distribution coefficients (Kd`s) on two soil types and a cement, using two different groundwater types. This study confirms that C-14 concentrations are significantly higher in the primary coolant from PWR stations compared to BWR stations. The C-14 followed trends of Co-60 generation during primary coolant demineralization at all but one of the stations examined. However, the C-14/Co-60 activity ratios measured by this study in resin samples through which samples of coolant were drawn were about 8 to 42 times higher than those reported for waste samples in the industry data base for PWR stations, and 15 to 730 times lower for the BWR stations.« less

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.

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.

Stress corrosion crack initiation of alloy 600 in PWR primary water

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

Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.

Stress corrosion crack (SCC) initiation of three mill-annealed (MA) alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular (IG) attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and IGSCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Processes controlling the SCC initiation in MA alloy 600 are discussed. INmore » PRESS, CORRECTED PROOF, 05/02/2017 - mfl« less

Posttest RELAP5 simulations of the Semiscale S-UT series experiments. [PWR

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

Leonard, M.T.

The RELAP5/MOD1 computer code was used to perform posttest calculations, simulating six experiments, run in the Semiscale Mod-2A facility, investigating the effects of upper head injection on small break transient behavior. The results of these calculations and corresponding test data are presented in this report. An evaluation is made of the capability of RELAP5 to calculate the thermal-hydraulic response of the Mod-2A system over a spectrum of break sizes, with and without the use of upper head injection.

Evaluation of Potential Locations for Siting Small Modular Reactors near Federal Energy Clusters to Support Federal Clean Energy Goals

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

Belles, Randy J.; Omitaomu, Olufemi A.

2014-09-01

Geographic information systems (GIS) technology was applied to analyze federal energy demand across the contiguous US. Several federal energy clusters were previously identified, including Hampton Roads, Virginia, which was subsequently studied in detail. This study provides an analysis of three additional diverse federal energy clusters. The analysis shows that there are potential sites in various federal energy clusters that could be evaluated further for placement of an integral pressurized-water reactor (iPWR) to support meeting federal clean energy goals.

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.

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

PubMed Central

2016-01-01

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

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

PubMed

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

2016-01-01

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

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

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

Sonat Sen; Gilles Youinou

2013-02-01

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

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

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

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

2013-07-01

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

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

PubMed

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

2012-02-01

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

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

NASA Astrophysics Data System (ADS)

Tuttle, James R.

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

Recent operating experiences with steam generators in Japanese NPPs

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

Yashima, Seiji

1997-02-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-06-01

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

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.

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.

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.

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

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

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

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

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

Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

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

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

2012-07-01

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

Limited Artificial and Natural Icing Tests Production UH-60A Helicopter (Re-Evaluation).

DTIC Science & Technology

1981-08-01

parameters , and definitions of icing types and severities are presented in appendix D. 2 RESULTS AND DISCUSSION GENERAL 9. Artificial and natural icing flight...anti-ice off, the system may be reactivated by cycling the appropriate windshield anti-ice switch. The windshield anti-ice system is fully operational...is off, then the fault monitor illuminates the respective PWR light on its front panel. The light informs the crew that further action is requied to

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

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...

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 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

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

Plasmon Spectroscopy Applied to Biomolecular Interactions in Membranes

NASA Astrophysics Data System (ADS)

Tollin, Gordon

2010-03-01

Plasmon-waveguide resonance (PWR) is an optical spectroscopy method that can provide information about materials immobilized on the surface of a plasmon resonator consisting of a right angle prism coated with thin layers of a metal (approx. 50 nm; usually silver) and a dielectric (approx. 500 nm; usually silica). The technique has been developed in our laboratory and is an extension of the more commonly used surface plasmon resonance (SPR) method, having higher sensitivity (20-50 fold) and resolution (10-20 fold). The dielectric layer allows plasmon excitation by light whose electric vector is polarized both perpendicular and parallel to the sensor surface, in contrast to SPR that can only utilize perpendicular polarized excitation. This allows both mass density and mass distribution to be characterized in uniaxially oriented deposited materials, such as biomembranes. We have utilized this technique to investigate binding interactions between membrane-incorporated protein receptors and their ligands (both proteins and small molecules), using both purified receptors inserted into lipid bilayers and membranes derived from cells expressing these receptors. Such studies have provided many new insights into biological signaling events. Inasmuch as many of these receptors are targets for approximately 50 percent of ethical drugs, PWR can be a useful methodology for drug discovery in the pharmaceutical industry. Examples of these experiments will be presented.

Determination of tube-to-tube support interaction characteristics. [PWR

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

Haslinger, K.H.

Tube-to-tube support interaction characteristics were determined on a multi-span tube geometry representative of the hot-leg side of the C-E, System 80 steam generator design. Results will become input for an autoclave type wear test program on steam generator tubes, performed by Kraftwerk Union (KWU). Correlation of test data reported here with similar data obtained from the wear tests will be performed in an attempt to make predictions about the long-term fretting behavior of steam generator tubes.

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

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

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

2012-07-01

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

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)

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

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

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

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

2013-07-01

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

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

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

Shaver, Mark W.; Lanning, Donald D.

2010-02-01

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

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

NASA Astrophysics Data System (ADS)

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

2014-02-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Shuttle Engine Designs Revolutionize Solar Power

NASA Technical Reports Server (NTRS)

2014-01-01

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

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

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

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

2014-12-16

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

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

PubMed

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

2014-12-16

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

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

Schulz, H.

In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A =more » open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.« less

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.

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

Development of new UV-I. I. Cerenkov Viewing Device

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

Kuribara, Masayuki; Nemoto, Koshichi

1994-02-01

The Cerenkov glow images from boiling-water reactors (BWR) and pressurized-water reactors (PWR) irradiated fuel assemblies are generally used for inspections. However, sometimes it is difficult or impossible to identify the image by the conventional Cerenkov Viewing Device (CVD), because of the long cooling time and/or low burnup. Now a new UV-I.I. (Ultra-Violet light Image Intensifier) CVD has been developed, which can detect the very weak Cerenkov glow from spent fuel assemblies. As this new device uses the newly developed proximity focused type UV-I.I., Cerenkov photons are used efficiently, producing better quality Cerenkov glow images. Moreover, since the image is convertedmore » to a video signal, it is easy to improve the signal to noise ratio (S/N) by an image processor. The new CVD was tested at BWR and PWR power plants in Japan, with fuel burnups ranging from 6,200--33,000 MWD/MTU (megawatt days per metric ton of uranium) and cooling times ranging from 370 to 6,200 d. The tests showed that the new CVD is superior to the conventional STA/CRIEPI CVD, and could detect very feeble Cerenkov glow images using an image processor.« less

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.

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

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

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

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

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

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

Nishizawa, Eiichi

1995-03-01

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

Comparison of Measures of Vibration Affecting Occupants of Military Vehicles

DTIC Science & Technology

1986-12-01

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

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

NASA Astrophysics Data System (ADS)

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

1986-10-01

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

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

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

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

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

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

J.W. Davis

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

Astronaut Robinson presents 2010 Silver Snoopy awards

NASA Image and Video Library

2010-06-23

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

Pretest mediction of Semiscale Test S-07-10 B. [PWR

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

Dobbe, C A

A best estimate prediction of Semiscale Test S-07-10B was performed at INEL by EG and G Idaho as part of the RELAP4/MOD6 code assessment effort and as the Nuclear Regulatory Commission pretest calculation for the Small Break Experiment. The RELAP4/MOD6 Update 4 and the RELAP4/MOD7 computer codes were used to analyze Semiscale Test S-07-10B, a 10% communicative cold leg break experiment. The Semiscale Mod-3 system utilized an electrially heated simulated core operating at a power level of 1.94 MW. The initial system pressure and temperature in the upper plenum was 2276 psia and 604/sup 0/F, respectively.

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

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.

TRIGA MARK-II source term

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

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

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.

Extension of the Bgl Broad Group Cross Section Library

NASA Astrophysics Data System (ADS)

Kirilova, Desislava; Belousov, Sergey; Ilieva, Krassimira

2009-08-01

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

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

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.

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)

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

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

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

2016-05-01

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

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

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

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

2016-05-12

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

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

NASA Astrophysics Data System (ADS)

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

2015-01-01

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

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

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

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

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

In-reactor performance of LWR-type tritium target rods

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

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

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

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

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

Zhang, Hongbin; Szilard, Ronaldo; Epiney, Aaron

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

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

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

Parameter study on the influence of prepressurization on PWR fuel rod behavior during normal operation and hypothetical LOCAs

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

Brzoska, B.; Depisch, F.; Fuchs, H.P.

To analyze the influence of prepressurization on fuel rod behavior, a parametric study has been performed that considers the effects of as-fabricated fuel rod internal prepressure on the normal operation and postulated loss-of-coolant accident (LOCA) rod behavior of a 1300-MW(electric) Kraftwerk Union (KWU) standard pressurized water reactor nuclear power plant. A variation of the prepressure in the range from 15 to 35 bars has only a slight influence on normal operation behavior. Considering the LOCA behavior, only a small temperature increase results from prepressure reduction, while the core-wide straining behavior is improved significantly. The KWU prepressurization takes both conditions intomore » account.« less

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

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

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

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

Faidy, C.

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

Planar Monolithic Schottky Varactor Diode Millimeter-Wave Frequency Multipliers

DTIC Science & Technology

1992-06-01

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

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

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

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

2016-08-30

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

Task related doses in Spanish pressurized water reactors over the period 1988-1992

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

O`Donnell, P.; Labarta, T.; Amor, I.

1995-03-01

In order to evaluate in depth the collective dose trend and its correlation with the effectiveness of the practical application of the ALARA principle in Spanish nuclear facilities, and base the different policy lines to promote this criteria, the CSN has fullfilled an analysis of the task related doses data over the period 1988-1992. Previously, the CSN had required to the utilities the compilation of their refuelling outage collective dose from 1988 according with a predeterminate number of tasks, in order to have available a representative and retrospective set of data in an homogeneous way and coherent with the internationalmore » data banks on occupational exposure in NPP, as the CEC and the NEA ones. The scope of this analysis was the following: first, the collective dose summaries for outage tasks and departments for PWR and for BWR, including the minimum, maximum and average dose (and statistics data) for 18 different refuelling outage tasks and 12 personal departments for each generation of each type of rector, the task and department related collective dose trends in each plant and in each generation, and second, the dose reduction techniques having been used during that period in each plant and the relative level of adoption. In this presentation the main results and conclusions of the first part of the study are reviewed for PWR.« less

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

Downar, Thomas

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISONmore » / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.« less

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.

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

PubMed

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

2012-06-06

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

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

NASA Astrophysics Data System (ADS)

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

2017-04-01

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

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

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

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

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

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

Vidal, Jean-Marc; Eschbach, Romain; Launay, Agnes

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of coolingmore » time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the CEA reference calculation codes for neutron physics with the JEFF-3.1.1 nuclear data set. (authors)« less

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

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.

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

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.

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

Reactor shutdown experience

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

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

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

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

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

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

1992-05-01

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

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

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

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

2013-07-01

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

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

DOE PAGES

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

2016-09-07

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

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

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

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

Kruska, Karen; Zhai, Ziqing; Bruemmer, Stephen M.

Due to its superior resistance to corrosion and stress corrosion cracking (SCC), high Cr, Ni-base Alloy 690 is now commonly used in pressurized water reactors (PWRs). Even though highly cold-worked (CW) Alloy 690 has been shown to be susceptible to SCC crack growth in PWR primary water environments, an open question remains whether SCC initiation was possible for these materials under constant load test conditions. Testing has been performed on a series of CW alloy 690 CRDM tubing specimens at constant load for up to 9,220 hours in 360°C simulated PWR primary water. A companion paper will discuss the overallmore » testing approach and describe results on different alloy 690 heats and cold work levels. The focus of the current paper is to illustrate the use of focused ion beam (FIB), scanning electron microscopy (SEM) and transmission electron microscopy (TEM) for the high-resolution investigation of precursor damage and intergranular (IG) crack nucleation in these specimens. Three-dimensional (3D) FIB/SEM imaging has been conducted on a series of grain boundary (GB) damage precursors, such as IG small cavities, local corrosion and even shallow cracks observed at the specimen surface. Contrast variations and EDS mapping were used to distinguish oxides, carbides and cavities from the matrix material. Nanometer-sized cavities were observed associated with GB carbides in the highly CW specimens. Shallow IG cracks were present in the 30%CW specimens and exhibited oxidized crack flanks and a higher density of cavities ahead of the oxide front in all cases. The shape and distribution of carbides and cavities in the plane of the cracked GBs was analyzed in 3D to gain a mechanistic understanding of the processes that may be leading to crack initiation in highly CW alloy 690.« less

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

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

NASA Astrophysics Data System (ADS)

Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

2014-08-01

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

Development and Application of Laser Peening System for PWR Power Plants

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

Masaki Yoda; Itaru Chida; Satoshi Okada

2006-07-01

Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion crackingmore » (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. (authors)« less

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.

Adsorbed radioactivity and radiographic imaging of surfaces of stainless steel and titanium

NASA Astrophysics Data System (ADS)

Jung, Haijo

1997-11-01

Type 304 stainless steel used for typical surface materials of spent fuel shipping casks and titanium were exposed in the spent fuel storage pool of a typical PWR power plant. Adsorption characteristics, effectiveness of decontamination by water cleaning and by electrocleaning, and swipe effectiveness on the metal surfaces were studied. A variety of environmental conditions had been manipulated to stimulate the potential 'weeping' phenomenon that often occurs with spent fuel shipping casks during transit. In a previous study, few heterogeneous effects of adsorbed contamination onto metal surfaces were observed. Radiographic images of cask surfaces were made in this study and showed clearly heterogeneous activity distributions. Acquired radiographic images were digitized and further analyzed with an image analysis computer package and compared to calibrated images by using standard sources. The measurements of activity distribution by using the radiographic image method were consistent with that using a HPGe detector. This radiographic image method was used to study the effects of electrocleaning for total and specified areas. The Modulation Transfer Function (MTF) of a film-screen system in contact with a radioactive metal surface was studied with neutron activated gold foils and showed more broad resolution properties than general diagnostic x-ray film-screen systems. Microstructure between normal areas and hot spots showed significant differences, and one hot spot appearing as a dot on the film image consisted of several small hot spots (about 10 μm in diameter). These hot spots were observed as structural defects of the metal surfaces.

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

Expected total counts for the Self-Interrogation Neutron Resonance Densitometry measurements of spent nuclear fuel

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

Rossa, Riccardo; Universite Libre de Bruxelles; Borella, Alessandro

2015-07-01

The Self-Interrogation Neutron Resonance Densitometry (SINRD) is a passive neutron technique that aims at a direct quantification of {sup 239}Pu in spent fuel assemblies by measuring the attenuation of the neutron flux in the energy region close to the 0.3 eV resonance of {sup 239}Pu. The {sup 239}Pu mass is estimated by calculating the SINRD signature, that is the ratio between the neutron counts in the fast energy region and around the 0.3 eV resonance region. The SINRD measurement approach in this study consisted in introducing a small neutron detector in the central guide tube of a PWR 17x17 fuelmore » assembly. In order to measure the neutron flux in the energy regions defined in the SINRD signature, different detector types were used. The response of a bare {sup 238}U fission chamber is considered for the determination of the fast neutron flux, while other thermal-epithermal detectors wrapped in neutron absorbers are envisaged to measure the neutron flux around the resonance region. This paper provides an estimation of the total neutron counts that can be achieved with the detector types proposed for the SINRD measurement. In the first section a set of detectors are evaluated in terms of total neutron counts and sensitivity to the {sup 239}Pu content, in order to identify the optimal measurement configuration for each detector type. Then a study is performed to increase the total neutron counts by increasing the detector size. The study shows that the highest total neutron counts are achieved by using either {sup 3}He or {sup 10}B proportional counters because of the high neutron efficiency of these detectors. However, the calculations indicate that the biggest contribution to the measurement uncertainty is due to the measurement of the fast neutron flux. Finally, similar sensitivity to the {sup 239}Pu content is obtained by using the different detector types for the measurement of the neutron flux close to the resonance region. Therefore, the total neutron counts associated to each detector type will play a major role in the selection of the detector types used for the SINRD measurement. (authors)« less

Analysis of steam generator tube rupture transients with single failure

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

Trambauer, K.

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

SCC Initiation Behavior of Alloy 182 in PWR Primary Water

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

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

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

Simultaneous optimization of loading pattern and burnable poison placement for PWRs

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

Alim, F.; Ivanov, K.; Yilmaz, S.

2006-07-01

To solve in-core fuel management optimization problem, GARCO-PSU (Genetic Algorithm Reactor Core Optimization - Pennsylvania State Univ.) is developed. This code is applicable for all types and geometry of PWR core structures with unlimited number of fuel assembly (FA) types in the inventory. For this reason an innovative genetic algorithm is developed with modifying the classical representation of the genotype. In-core fuel management heuristic rules are introduced into GARCO. The core re-load design optimization has two parts, loading pattern (LP) optimization and burnable poison (BP) placement optimization. These parts depend on each other, but it is difficult to solve themore » combined problem due to its large size. Separating the problem into two parts provides a practical way to solve the problem. However, the result of this method does not reflect the real optimal solution. GARCO-PSU achieves to solve LP optimization and BP placement optimization simultaneously in an efficient manner. (authors)« less

76 FR 41783 - Combined Notice of Filings #2

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-15

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

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

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

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

2012-07-01

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

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

Samohyl, P.

The application of the LBB requires also fatigue flaw growth assessment. This analysis was performed for PWR nuclear power plants types VVER 440/230, VVER 440/213c, VVER 1000/320. Respecting that these NPP`s were designed according to Russian codes that differ from US codes it was needed to compare these approaches. Comparison with our experimental data was accomplished, too. Margins of applicability of the US methods and their modifications for the materials used for construction of Czech and Slovak NPP`s are shown. Computer code accomplishing the analysis according to described method is presented. Some measurement and calculations show that thermal stratifications inmore » horizontal pipelines can lead to additive loads that are not negligible and can be dangerous. An attempt to include these loads induced by steady-state stratification was made.« less

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

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

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

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

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

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

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

Zodiates, A.M.; Willcock, A.

1995-03-01

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

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

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

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

2006-07-01

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

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

NASA Astrophysics Data System (ADS)

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

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

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

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

Shen, W.

2012-07-01

Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, threemore » benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)« less

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

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

None

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

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

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

Walter, Matthew; Yin, Shengjun; Stevens, Gary

2012-01-01

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

Postpartum weight trajectories in overweight and lean women.

PubMed

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

2017-06-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-07-01

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

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

PubMed

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

2015-01-01

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

Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

DTIC Science & Technology

2013-06-01

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

Accident analysis of heavy water cooled thorium breeder reactor

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

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

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

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

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

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

This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal-mechanical stress analysis were performed for typical thermal-mechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, 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 inmore » US-NRC report: NUREG-6909.« less

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

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

Gilles Youinou; Andrea Alfonsi

2012-03-01

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

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

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

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

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

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

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

Hirofumi Okabe; Tatsuaki Sato; Yuichi Shoji

2013-07-01

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

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

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

NASA Astrophysics Data System (ADS)

Sawicki, Jerzy A.

2011-08-01

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

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

NASA Astrophysics Data System (ADS)

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

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

Development of burnup dependent fuel rod model in COBRA-TF

NASA Astrophysics Data System (ADS)

Yilmaz, Mine Ozdemir

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

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-25

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

Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

DOEpatents

Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

1980-06-06

An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

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

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

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

1995-09-01

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

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

NASA Technical Reports Server (NTRS)

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

2008-01-01

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

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

DOE PAGES

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

2015-05-17

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

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

PubMed

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

2018-05-29

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

PWR steam generator chemical cleaning, Phase I. Final report

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

Rothstein, S.

1978-07-01

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

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

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

None

1979-10-01

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

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

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

Hsu, M.T.; Davis, C.B.; Behling, S.R.

This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio wasmore » maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).« 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.

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

PubMed Central

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

2016-01-01

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

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

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

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

2011-07-01

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

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

PubMed Central

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

2013-01-01

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

EMERALD REV.1. PWR Accident Activity Release

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

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

1975-10-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-10-01

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

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

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.

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

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

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

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

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

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

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

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

Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Dufek, Jan

2014-06-01

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

Commercial high-level-waste management: Options and economics. A comparative analysis of the ceramic and glass waste forms

NASA Astrophysics Data System (ADS)

McKisson, R. L.; Grantham, L. F.; Guon, J.; Recht, H. L.

1983-02-01

Results of an estimate of the waste management costs of the commercial high level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20 year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. Waste loading and waste form density are the driving factors in that the low waste loading (25%) and relatively low density (3.1 g cu cm) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 cu cm.

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

NASA Astrophysics Data System (ADS)

Burdo, James S.

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

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

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

PubMed

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

2018-05-19

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

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.

Immunogenicity of a recombinant fusion protein of tandem repeat epitopes of foot-and-mouth disease virus type Asia 1 for guinea pigs.

PubMed

Zhang, Q; Yang, Y Q; Zhang, Z Y; Li, L; Yan, W Y; Jiang, W J; Xin, A G; Lei, C X; Zheng, Z X

2002-01-01

In this study, the sequences of capsid protein VPI regions of YNAs1.1 and YNAs1.2 isolates of foot-and-mouth disease virus (FMDV) were analyzed and a peptide containing amino acids (aa) 133-158 of VP1 and aa 20-34 of VP4 of FMDV type Asia I was assumed to contain B and T cell epitopes, because it is hypervariable and includes a cell attachment site RGD located in the G-H loop. The DNA fragments encoding aa 133-158 of VP1 and aa 20-34 of VP4 of FMDV type Asia 1 were chemically synthesized and ligated into a tandem repeat of aa 133-158-20 approximately 34-133-158. In order to enhance its immunogenicity, the tandem repeat was inserted downstream of the beta-galactosidase gene in the expression vector pWR590. This insertion yielded a recombinant expression vector pAS1 encoding the fusion protein. The latter reacted with sera from FMDV type Asia 1-infected animals in vitro and elicited high levels of neutralizing antibodies in guinea pigs. The T cell proliferation in immunized animals increased following stimulation with the fusion protein. It is reported for the first time that a recombinant fusion protein vaccine was produced using B and T cell epitopes of FMDV type Asia 1 and that this fusion protein was immunogenic. The fusion protein reported here can serve as a candidate of fusion epitopes for design of a vaccine against FMDV type Asia 1.

A study of life prediction differences for a nickel-base Alloy 690 using a threshold and a non-threshold model

NASA Astrophysics Data System (ADS)

Young, B. A.; Gao, Xiaosheng; Srivatsan, T. S.

2009-10-01

In this paper we compare and contrast the crack growth rate of a nickel-base superalloy (Alloy 690) in the Pressurized Water Reactor (PWR) environment. Over the last few years, a preponderance of test data has been gathered on both Alloy 690 thick plate and Alloy 690 tubing. The original model, essentially based on a small data set for thick plate, compensated for temperature, load ratio and stress-intensity range but did not compensate for the fatigue threshold of the material. As additional test data on both plate and tube product became available the model was gradually revised to account for threshold properties. Both the original and revised models generated acceptable results for data that were above 1 × 10 -11 m/s. However, the test data at the lower growth rates were over-predicted by the non-threshold model. Since the original model did not take the fatigue threshold into account, this model predicted no operating stress below which the material would effectively undergo fatigue crack growth. Because of an over-prediction of the growth rate below 1 × 10 -11 m/s, due to a combination of low stress, small crack size and long rise-time, the model in general leads to an under-prediction of the total available life of the components.

78 FR 35960 - Minor Boundary Revision at Mojave National Preserve

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-14

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

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

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

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

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

Reactor antineutrino detector iDREAM.

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Regeneratively Cooled Liquid Oxygen/Methane Technology Development

NASA Technical Reports Server (NTRS)

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

2012-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-12-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

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

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

Van Der Sluys, W.A.

1982-12-01

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

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

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

NASA Astrophysics Data System (ADS)

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

1994-02-01

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

Improved Biomolecular Thin-Film Sensor based on Plasmon Waveguide Resonance

NASA Astrophysics Data System (ADS)

Byard, Courtney; Aslan, Mustafa; Mendes, Sergio

2009-05-01

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

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

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

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

1983-04-01

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

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

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

Li, Jie; Liu, Yung Y.

2016-05-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-05-01

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

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

PubMed Central

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

2015-01-01

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

Derivation of the Korean radwaste scaling factor

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

Kwang Yong Jee; Hong Joo Ahn; Se Chul Sohn

2007-07-01

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

Current forgings and their properties for steam generator of nuclear plant

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

Tsukada, Hisashi; Suzuki, Komei; Kusuhashi, Mikio

1997-12-31

Current steel forgings for steam generator (SG) of PWR plant are reviewed in the aspect of design and material improvement. The following three items are introduced. The use of integral type steel forgings for the fabrication of steam generator enhances the structural integrity and makes easier fabrication and inspection including in-service inspection. The following examples of current integral type forgings developed by the Japan Steel Works, Ltd. (JSW) are introduced: (1) primary head integrated with nozzles, manways and supports; (2) steam drum head integrated with nozzle and handhole; (3) conical shell integrated with cylindrical sections and handholes. In order tomore » decrease the weight of steam generator, the high strength materials such as SA508, Cl.3a steel have been adopted in some cases. The properties of this steel are introduced and the chemistry and heat treatment condition are discussed. As one of the methods to minimize the macro- and micro-segregations, the use of vacuum carbon deoxidation (VCD), i.e. deoxidization of steel by gaseous CO reaction, with addition of Al for grain refining was investigated. The properties of SA508, Cl.3 steels with Low Si content are compared with those of conventional one.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-20

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-19

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

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

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

None

2002-11-01

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

75 FR 15752 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-30

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

75 FR 18907 - Sunshine Federal Register Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-13

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

Microchemical and microstructural evolution of AISI 304 stainless steel irradiated in EBR-II at PWR-relevant dpa rates

NASA Astrophysics Data System (ADS)

Dong, Y.; Sencer, B. H.; Garner, F. A.; Marquis, E. A.

2015-12-01

AISI 304 stainless steel was irradiated at 416 °C and 450 °C at a 4.4 × 10-9 and 3.05 × 10-7 dpa/s to ∼0.4 and ∼28 dpa, respectively, in the reflector of the EBR-II fast reactor. Both unirradiated and irradiated conditions were examined using standard and scanning transmission electron microscopy, energy dispersive spectroscopy, and atom probe tomography on very small specimens produced by focused ion beam milling. These results are compared with previous electron microscopy examination of 3 mm disks from essentially the same material. By comparing a very low dose specimen with a much higher dose specimen, both derived from a single reactor assembly, it has been demonstrated that the coupled microstructural and microchemical evolution of dislocation loops and other sinks begins very early, with elemental segregation producing at these sinks what appears to be measurable precursors to fully formed precipitates found at higher doses. The nature of these sinks and their possible precursors are examined in detail.

Final Prep on SSME

NASA Image and Video Library

2005-10-25

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

Final Prep on SSME

NASA Technical Reports Server (NTRS)

2005-01-01

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

DEVELOPMENT OF WELDED SEAL FOR S3G REACTOR VESSEL

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

Rogers, J.W.

1958-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-12

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-19

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-11

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

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

ERIC Educational Resources Information Center

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

2013-01-01

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

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

DOEpatents

Tokarz, R.D.

1981-10-27

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

78 FR 44596 - Minor Boundary Revision at Yosemite National Park

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-24

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-16

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

Counterfeit Parts Prevention Strategy Guide Product Overview

DTIC Science & Technology

2014-05-08

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

Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

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

Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

2013-07-01

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-22

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

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

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

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

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

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

Code of Federal Regulations, 2012 CFR

2012-07-01

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

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

Code of Federal Regulations, 2013 CFR

2013-07-01

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

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

Code of Federal Regulations, 2014 CFR

2014-07-01

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

75 FR 13800 - Sunshine Federal Register Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-23

..., Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance (Public Meeting..., 2010. Week of April 26, 2010--Tentative Thursday, April 29, 2010 9:30 a.m. Briefing on the Fuel Cycle... and Work Life Branch, at 301-492-2230, TDD: 301-415-2100, or by e-mail at [email protected

77 FR 65906 - Minor Boundary Revision at Minidoka National Historic Site

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-31

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-MIIN-11234; 9360-726] Minor Boundary Revision at Minidoka National Historic Site AGENCY: National Park Service, Interior. ACTION: Notification of boundary revision. SUMMARY: Notice is hereby given that, pursuant to 16 U.S.C. 460l- 9(c)(1)(ii...

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

76 FR 24514 - Honouliuli Special Resource Study, Honolulu, Maui, Hawaii, and Kauai Counties, HI

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-02

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-0308-6923;9082-HONO-420] Honouliuli.... Background: As authorized by the Department of the Interior, Environment, and Related Agencies Appropriations... State of Hawaii with respect to (1) Their significance as components of World War II; (2) significance...

76 FR 22917 - Dog Management Plan/Draft Environmental Impact Statement, Golden Gate National Recreation Area...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-25

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO--0315-696; 8145-8B90-SZM] Dog... Impact Statement/Dog Management Plan, Golden Gate National Recreation Area. SUMMARY: The National Park Service has prepared a Draft Dog Management Plan and Environmental Impact Statement (Plan/DEIS). The Plan...

Amorphous and Nanocrystalline High Temperature Magnetic Material for PWR

DTIC Science & Technology

2006-03-01

FOR PUBLICATION. //Signature// //Signature// ______________________________________ __________________________________ JOHN C ...times that of conventional ferrites at room temperature); 2) Frequency: 200 kHz to 1 MHz; 3) Temperature: 200 ° C and above. The goals of the DUST...NAME OF RESPONSIBLE PERSON (Monitor) a. REPORT Unclassified b. ABSTRACT Unclassified c . THIS PAGE Unclassified 17. LIMITATION OF ABSTRACT

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

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

NASA Astrophysics Data System (ADS)

Tadesse, Abel; Fredriksson, Hasse

2018-06-01

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

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

Overman, Nicole R.; Toloczko, Mychailo B.; Olszta, Matthew J.

High chromium, nickel-base Alloy 690 exhibits an increased resistance to stress corrosion cracking (SCC) in pressurized water reactor (PWR) primary water environments over lower chromium alloy 600. As a result, Alloy 690 has been used to replace Alloy 600 for steam generator tubing, reactor pressure vessel nozzles and other pressure boundary components. However, recent laboratory crack-growth testing has revealed that heavily cold-worked Alloy 690 materials can become susceptible to SCC. To evaluate reasons for this increased SCC susceptibility, detailed characterizations have been performed on as-received and cold-worked Alloy 690 materials using electron backscatter diffraction (EBSD) and Vickers hardness measurements. Examinationsmore » were performed on cross sections of compact tension specimens that were used for SCC crack growth rate testing in simulated PWR primary water. Hardness and the EBSD integrated misorientation density could both be related to the degree of cold work for materials of similar grain size. However, a microstructural dependence was observed for strain correlations using EBSD and hardness which should be considered if this technique is to be used for gaining insight on SCC growth rates« less

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

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

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

2006-07-01

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

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

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

Parisi, Carlo; Prescott, Steve; Ma, Zhegang

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

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-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

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

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

J-2X Turbopump Cavitation Diagnostics

NASA Technical Reports Server (NTRS)

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

2010-01-01

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

User Data Package - Energy-Efficient Windows and Window Coverings for Naval Housing

DTIC Science & Technology

1990-07-01

1765 33 Savannah 1819 32 Tucson 1800 32 Winslow 4782 35 Idaho Yuma 974 33 Boise 5809 44 Lewiston 5542 46 Arkansas Pocatello 7033 43 Fort Smith 3292 35...Alexandria, VA DODDS Pac, FAC, Okinawa, Japan DOE Fed Energy Mgt Program, Wash, DC: INEL Tech Lib Reports Sta. Idaho Falls. ID; Knolls Atomic Pwr Lab

77 FR 9960 - Final Environmental Impact Statement for Extension of F-Line Streetcar Service to Fort Mason...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-21

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-1103-8840; 2051-P580-579] Final... AGENCY: National Park Service, Department of the Interior. ACTION: Notice of Availability of the Final... resources. Many also suggested various design ideas and other measures to help reduce these impacts. In...

78 FR 38359 - Approval of Record of Decision for Relocation of Cattle Point Road, San Juan Island National...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-26

... DEPARTMENT OF THE INTERIOR National Park Service [NPS-PWR-PWRO-12863; PPPWSAJHA0 PPMPSAS1Z.Y00000... Park, San Juan County, Washington AGENCY: National Park Service, Interior. ACTION: Notice of Record of....2), the Department of the Interior, National Park Service has prepared and approved a Record of...

77 FR 31655 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-29

... controverted. In addition, the requestor/petitioner shall provide a brief explanation of the bases for the... against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR [Pressurized... Institute] 97-06, Revision 3, ``Steam Generator Program Guidelines'' (Reference 1) and RG 1.121, ``Bases for...

Manufacturing Methods and Technology Measure for Fabrication of Silicon Transcalent Rectifier.

DTIC Science & Technology

1980-09-01

Prod Test/Eval’, z HA Kotler a Patent- Power & E 1 RM Roderick Env. Eng. & Test 1 JB Grosh Iron Mouptain - .l TUBE PARTS MFG. 5 RL SPALDING...AFAL/PODI ATTN: Working Group on Pwr. Devices (Mr. Philip Herron) 201 Varick Street Wright Patterson AFB, OH 45433 New York, NY 10014 Commander Mr

Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

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

Chen, Y.; Alexandreanu, B.; Natesan, K.

2014-11-01

Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3more » were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.« less

High-Resolution Characterizations of Grain Boundary Damage and Stress Corrosion Cracks in Cold-Rolled Alloy 690

NASA Astrophysics Data System (ADS)

Bruemmer, S. M.; Olszta, M. J.; Toloczko, M. B.; Thomas, L. E.

Unidirectional cold rolling has been shown to promote intergranular stress corrosion cracking (IGSCC) in alloy 690 tested in PWR primary water. High-resolution scanning (SEM) and transmission electron microscopy (TEM) have been employed to investigate the microstructural reasons for this enhanced susceptibility in two stages, first examining grain boundary damage produced by cold rolling and second by characterization of stress corrosion crack tips. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG precipitate distribution. 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. For the same degree of cold rolling, alloys with few IG carbides exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal no interaction between the preexisting voids and cracked carbides with the propagation. In many cases, these features appeared to blunt propagation of IGSCC cracks. High-resolution characterizations are described for cold-rolled alloy 690 CRDM tubing and plate materials to gain insights into IGSCC mechanisms.

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

NASA Astrophysics Data System (ADS)

Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

2013-09-01

Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

FY2012 summary of tasks completed on PROTEUS-thermal work.

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

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

76 FR 23577 - Combined Notice of Filings #1

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-27

... Interconnection, L.L.C. submits tariff filing per 35.13(a)(2)(iii: Queue No. W3-124--Original Service Agreement No... Transmission Agreement with Auburndale Pwr Partners to be effective 5/1/2011. Filed Date: 04/21/2011. Accession... tariff filing per 35.13(a)(1): 04--21--11 Paris Rate Schedule 407 Settlement to be effective 3/31/2011...

Steam generators regulatory practices and issues in Spain

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

Mendoza, C.; Castelao, C.; Ruiz-Colino, J.

1997-02-01

This paper presents the actual status of Spanish Steam Generator tubes, actions developed by PWR plant owners and submitted to CSN, and regulatory activities related to tube degradation mechanisms analysis; NDT tube inspection techniques; tube, tubesheet and TSPs integrity studies; tube plugging/repair criteria; preventive and corrective measures including whole SGs replacement; tube leak measurement methods and other operational aspects.

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

ERIC Educational Resources Information Center

Reihman, Thomas C.

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

Organometallics in High Energy Chemistry.

DTIC Science & Technology

1983-10-31

Luines Physeical ftaenc Chemistry DepatneWu. SJI International. Menlo PWr *. CaiOwrnia M10 Rceived Nouvber 8. 1IM The otslytic formation of6nw carbon...support the idea that the metalloazocyclopropane intermediate is the reactive intermediate that leads to transalkylation. A discussion of the...exceptionally good correlation between the catalytic reactivity patterns of palladium black in its reactions with tertiary amines and those of homogeneous

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

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

Crack opening area estimates in pressurized through-wall cracked elbows under bending

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

Franco, C.; Gilles, P.; Pignol, M.

1997-04-01

One of the most important aspects in the leak-before-break approach is the estimation of the crack opening area corresponding to potential through-wall cracks at critical locations during plant operation. In order to provide a reasonable lower bound to the leak area under such loading conditions, numerous experimental and numerical programs have been developed in USA, U.K. and FRG and widely discussed in literature. This paper aims to extend these investigations on a class of pipe elbows characteristic of PWR main coolant piping. The paper is divided in three main parts. First, a new simplified estimation scheme for leakage area ismore » described, based on the reference stress method. This approach mainly developed in U.K. and more recently in France provides a convenient way to account for the non-linear behavior of the material. Second, the method is carried out for circumferential through-wall cracks located in PWR elbows subjected to internal pressure. Finite element crack area results are presented and comparisons are made with our predictions. Finally, in the third part, the discussion is extended to elbows under combined pressure and in plane bending moment.« less

Joining dissimilar stainless steels for pressure vessel components

NASA Astrophysics Data System (ADS)

Sun, Zheng; Han, Huai-Yue

1994-03-01

A series of studies was carried out to examine the weldability and properties of dissimilar steel joints between martensitic and austenitic stainless steels - F6NM (OCr13Ni4Mo) and AISI 347, respectively. Such joints are important parts in, e.g. the primary circuit of a pressurized water reactor (PWR). This kind of joint requires both good mechanical properties, corrosion resistance and a stable magnetic permeability besides good weldability. The weldability tests included weld thermal simulation of the martensitic steel for investigating the influence of weld thermal cycles and post-weld heat treatment (PWHT) on the mechanical properties of the heat-affected zone (HAZ); implant testing for examining the tendency for cold cracking of martensitic steel; rigid restraint testing for determining hot crack susceptibility of the multi-pass dissimilar steel joints. The joints were subjected to various mechanical tests including a tensile test, bending test and impact test at various temperatures, as well as slow strain-rate test for examining the stress corrosion cracking tendency in the simulated environment of a primary circuit of a PWR. The results of various tests indicated that the quality of the tube/tube joints is satisfactory for meeting all the design requirements.

Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

NASA Astrophysics Data System (ADS)

Chen, Junjie; Xiao, Qian; Lu, Zhanpeng; Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan

2017-06-01

The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters.

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

On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

PubMed

Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

2016-08-01

For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

Annual progress report on the NSRR experiments, (21)

NASA Astrophysics Data System (ADS)

1992-05-01

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into pre-irradiated fuel tests and fresh fuel tests; the former includes 2 JMTR pre-irradiated fuel tests, 2 PWR pre-irradiated fuel tests, and 2 BWR pre-irradiated fuel tests, and the latter includes 6 standard fuel tests (6 SP(center dot)CP scoping tests), 7 power and cooling condition parameter tests (4 flow shrouded fuel tests, 1 bundle simulation test, 1 fully water-filled vessel test, 1 high pressure/high temperature loop test), 12 special fuel tests (3 stainless steel clad fuel tests, 3 improved PWR fuel tests, 6 improved BWR fuel tests), 3 severe fuel damage tests (1 high temperature flooding test, 1 flooding behavior observation test, 1 debris coolability test), 3 fast breeder reactor fuel tests (2 moderator material characteristic measurement tests, 1 fuel behavior observation test), and 2 miscellaneous tests (2 preliminary tests for pre-irradiated fuel tests).

"Small Steps, Big Rewards": Preventing Type 2 Diabetes

MedlinePlus

... please turn Javascript on. Feature: Diabetes "Small Steps, Big Rewards": Preventing Type 2 Diabetes Past Issues / Fall ... These are the plain facts in "Small Steps. Big Rewards: Prevent Type 2 Diabetes," an education campaign ...

10 CFR 50.75 - Reporting and recordkeeping for decommissioning planning.

Code of Federal Regulations, 2010 CFR

2010-01-01

... investing or otherwise, that a prudent investor would use in the same circumstances. The term “prudent... than or equal to 3400 MWt $105 between 1200 MWt and 3400 MWt (For a PWR of less than 1200 MWt, use P... 3400 MWt (For a BWR of less than 1200 MWt, use P=1200 MWt) $(104+0.009P) (2) An adjustment factor at...

Experimental and Analytical Development of the Application of a Transit Laser Velocimeter

DTIC Science & Technology

1980-11-01

from 0.005" brass shim stock with carefully finished edges and chemically blackened surface and is slightly adjustable in position to compensate for...personnel by Mr. T. V. C i e l , ETF . Mr. V i r g i l Cline, PWr, and t h e f a c i l i t y technic ians . A 1" diameter underexpanded unheated

40 CFR 59.507 - What are the labeling requirements for aerosol coatings?

Code of Federal Regulations, 2010 CFR

2010-07-01

... in Table 1 of this subpart or a company-specific code, if that code is explained as required by § 59.511(a); (2) The applicable PWR limit for the product specified in Table 1 of this subpart; (3) The day... this subpart. (b) The label on the product must be displayed in such a manner that it is readily...

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

Development of Hplc Techniques for the Analysis of Trace Metal Species in the Primary Coolant of a Pressurised Water Reactor.

NASA Astrophysics Data System (ADS)

Barron, Keiron Robert Philip

Available from UMI in association with The British Library. The need to monitor corrosion products in the primary circuit of a pressurised water reactor (PWR), at a concentration of 10pg ml^{-1} is discussed. A review of trace and ultra-trace metal analysis, relevant to the specific requirements imposed by primary coolant chemistry, indicated that high performance liquid chromatography (HPLC), coupled with preconcentration of sample was an ideal technique. A HPLC system was developed to determine trace metal species in simulated PWR primary coolant. In order to achieve the desired detection limit an on-line preconcentration system had to be developed. Separations were performed on Aminex A9 and Benson BC-X10 analytical columns. Detection was by post column reaction with Eriochrome Black T and Calmagite Linear calibrations of 2.5-100ng of cobalt (the main species of interest), were achieved using up to 200ml samples. The detection limit for a 200ml sample was 10pg ml^{-1}. In order to achieve the desired aim of on-line collection of species at 300^circ C, the use of inorganic ion-exchangers is essential. A novel application, utilising the attractive features of the inorganic ion-exchangers titanium dioxide, zirconium dioxide, zirconium arsenophosphate and pore controlled glass beads, was developed for the preconcentration of trace metal species at temperature and pressure. The performance of these exchangers, at ambient and 300^ circC was assessed by their inclusion in the developed analytical system and by the use of radioisotopes. The particular emphasis during the development has been upon accuracy, reproducibility of recovery, stability of reagents and system contamination, studied by the use of radioisotopes and response to post column reagents. This study in conjunction with work carried out at Winfrith, resulted in a monitoring system that could follow changes in coolant chemistry, on deposition and release of metal species in simulated PWR water loops. On -line detection of cobalt at 11pg ml^{ -1} was recorded, something which previously could not be performed by other techniques.

"Small Steps, Big Rewards": You Can Prevent Type 2 Diabetes

MedlinePlus

... Home Current Issue Past Issues Special Section "Small Steps, Big Rewards": You Can Prevent Type 2 Diabetes ... onset. Those are the basic facts of "Small Steps. Big Rewards: Prevent type 2 Diabetes," created by ...

Small Intestine Cancer—Health Professional Version

Cancer.gov

Adenocarcinoma is the most common type of small intestine cancer. Other types of small intestine cancer are sarcomas, carcinoid tumors, gastrointestinal stromal tumors, and lymphomas. Find evidence-based information on small intestine cancer treatment, research, and statistics.

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

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

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

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

Fusion Helmet: Electronic Analysis

DTIC Science & Technology

2014-04-01

Table 1: LYR203-101B Board Feature P1 (SEC MODULE) DM648 GPIO PORn Video Ports (2) Bootmode SPI/UART I2C CLKIN MDIO DDR2 128MB/16bit SPI Flash 16...McASP EMAC-SGMII /2 MDIO I2C GPIO DDR2 128MB/16bit JTAG Memory CLKGEN I2C PGoodPGood PORn Pwr LED Power DSP SPI/UART DSP SPI/UARTSPI/UART Video Display

PWR PRELIMINARY DESIGN FOR PL-3

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

Humphries, G. E.

1962-02-28

The pressurized water reactor preliminary design, the preferred design developed under Phase I of the PL-3 contract, is presented. Plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentatlon description, shielding description, auxiliary systems, power plant equipment, waste dispusal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information, and a training program outline are given. (auth)

Solar Photovoltaic and Liquid Natural Gas Opportunities for Command Naval Region Hawaii

DTIC Science & Technology

2014-12-01

Utilities Commission xii PV Photovoltaic Pwr Power RE Renewable Energy Re-gas Regasification RFP Request For Proposal RMI Rocky... forecasted LS diesel price and the forecasted LNG delivered-to-the- power -plant cost. The forecast for LS diesel by FGE from year 2020–2030 is seen...annual/html/epa_08_01.html Electric Power Research Institute. (July, 2010). Addressing solar photovoltaic operations and maintenance challenges: A

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

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

Stallmann, F.W.

1984-08-01

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

Two-phase Fluid Selection for High-temperature Automotive Platforms

DTIC Science & Technology

2012-09-01

cases, extra work by researchers can be done to uprate the device, either by parameter conformance, parameter re-characterization, or stress balancing...chemical reactivity , noble metal oxidization, intermetallic growth, CTE mismatch and other failure mechanisms are an issue with wide bandgap...21005-1852 1 US ARMY TARDEC ATTN AMSRD TAR E/ PWR C SPANGLER 6501 E 11 MILE RD, BLDG 212 WARREN MI 48397-5000 1 UNIV OF MARYLAND

CH-47C Vulnerability Reduction Modification Program - Fly-by-Wire Backup Demonstration

DTIC Science & Technology

1976-08-01

Actuator Position for Combined Axis Input ............................. 91 4 Systems Assessment Summary................... 95 C-1 Instrumentation Parameters ...SERVO CARD jEETO FROM MIXERS SUfEV __________ HYLIC AMPL AMPLVLE SHUT-O- DOWN DC PWR LOGIC REA MIONITOR SUMMER *O:EO SWITCH- BUFFER OVER 1 NETWORK...and ranels (Figures 12 and 13). The existing DELS preflight test set, which provides access to the system parameters , was installed along with the

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

NASA Astrophysics Data System (ADS)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

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

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

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

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

2012-07-01

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

Current training initiatives at Nuclear Electric plc

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

Fowler, C.D.

1993-01-01

Nuclear Electric, one of the three generating companies to emerge from the demise of the U.K.'s Central Electricity Generating Board (CEGB), owns and operates the commercial nuclear power stations in England and Wales. The U.K. government proscribed further construction beyond Sizewell B, the United Kingdom's first pressurized water reactor (PWR) station, pending the outcome of a review of the future of nuclear power to be held in 1994. The major challenges facing Nuclear Electric at its formation in 1990 were therefore to demonstrate that nuclear power is safe, economical, and environmentally acceptable and to complete the PWR station under constructionmore » on time and within budget. A significant number of activities were started that were designed to increase output, reduce costs, and ensure that the previous excellent safety standards were maintained. A major activity was to reduce the numbers of staff employed, with a recognition from the outset that this reduction could only be achieved with a significant human resource development program. Future company staff would have to be competent in more areas and more productive. This paper summarizes some of the initiatives currently being pursued throughout the company and the progress toward ensuring that staff with the required competences are available to commission and operate the Sizewell B program in 1994.« less

System Engineering for J-2X Development: The Simpler, the Better

NASA Technical Reports Server (NTRS)

Kelly, William M.; Greasley, Paul; Greene, William D.; Ackerman, Peter

2008-01-01

The Ares I and Ares V Vehicles will utilize the J-2X rocket engine developed for NASA by the Pratt and Whitney Rocketdyne Company (PWR) as the upper stage engine (USE). The J-2X is an improved higher power version of the original J-2 engine used for Apollo. System Engineering (SE) facilitates direct and open discussions of issues and problems. This simple idea is often overlooked in large, complex engineering development programs. Definition and distribution of requirements from the engine level to the component level is controlled by Allocation Reports which breaks down numerical design objectives (weight, reliability, etc.) into quanta goals for each component area. Linked databases of design and verification requirements help eliminate redundancy and potential mistakes inherent in separated systems. Another tool, the Architecture Design Description (ADD), is used to control J-2X system architecture and effectively communicate configuration changes to those involved in the design process. But the proof of an effective process is in successful program accomplishment. SE is the methodology being used to meet the challenge of completing J-2X engine certification 2 years ahead of any engine program ever developed at PWR. This paper describes the simple, better SE tools and techniques used to achieve this success.

Investigation into the effect of water chemistry on corrosion product formation in areas of accelerated flow

NASA Astrophysics Data System (ADS)

McGrady, John; Scenini, Fabio; Duff, Jonathan; Stevens, Nicholas; Cassineri, Stefano; Curioni, Michele; Banks, Andrew

2017-09-01

The deposition of CRUD (Chalk River Unidentified Deposit) in the primary circuit of a Pressurised Water Reactor (PWR) is known to preferentially occur in regions of the circuit where flow acceleration of coolant occurs. A micro-fluidic flow cell was used to recreate accelerated flow under simulated PWR conditions, by flowing water through a disc with a central micro-orifice. CRUD deposition was reproduced on the disc, and CRUD Build-Up Rates (BUR) in various regions of the disc were analysed. The effect of the local environment on BUR was investigated. In particular, the effect of flow velocity, specimen material and Fe concentration were considered. The morphology and composition of the deposits were analysed with respect to experimental conditions. The BUR of CRUD was found to be sensitive to flow velocity and Fe concentration, suggesting that mass transfer is an important factor. The morphology of the deposit was affected by the specimen material indicating a dependence on surface/particle electrostatics meaning surface chemistry plays an important role in deposition. The preferential deposition of CRUD in accelerated flow regions due to electrokinetic effects was observed and it was shown that higher Fe concentrations in solution increased BURs within the orifice whereas increased flow velocity reduced BURs.

Penetrative Internal Oxidation from Alloy 690 Surfaces and Stress Corrosion Crack Walls during Exposure to PWR Primary Water

NASA Astrophysics Data System (ADS)

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

Analytical electron microscopy and three-dimensional atom probe tomography (ATP) examinations of surface and near-surface oxidation have been performed on Ni-30%Cr alloy 690 materials after exposure to high-temperature, simulated PWR primary water. The oxidation nanostructures have been characterized at crack walls after stress-corrosion crack growth tests and at polished surfaces of unstressed specimens for the same alloys. Localized oxidation was discovered for both crack walls and surfaces as continuous filaments (typically <10 nm in diameter) extending from the water interface into the alloy 690 matrix reaching depths of 500 nm. These filaments consisted of discrete, plate-shaped Cr2O3 particles surrounded by a distribution of nanocrystalline, rock-salt (Ni-Cr-Fe) oxide. The oxide-containing filament depth was found to increase with exposure time and, at longer times, the filaments became very dense at the surface leaving only isolated islands of metal. Individual dislocations were oxidized in non-deformed materials, while the oxidation path appeared to be along more complex dislocation substructures in heavily deformed materials. This paper will highlight the use of high resolution scanning and transmission electron microscopy in combination with APT to better elucidate the microstructure and microchemistry of the filamentary oxidation.

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

NASA Technical Reports Server (NTRS)

Sargent, Scott R.; Becht, David G.

2011-01-01

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

Development Status of the CECE Cryogenic Deep Throttling Demonstrator Engine

NASA Technical Reports Server (NTRS)

2008-01-01

As one of the first technology development programs awarded by NASA under the U.S. Space Exploration Policy (USSEP), the Pratt & Whitney Rocketdyne (PWR) Deep Throttling, Common Extensible Cryogenic Engine (CECE) program was selected by NASA in November 2004 to begin technology development and demonstration toward a deep throttling, cryogenic engine supporting ongoing trade studies for NASA's Lunar Lander descent stage. The CECE program leverages the maturity and previous investment of a flight-proven hydrogen/oxygen expander cycle engine, the PWR RLI0, to develop and demonstrate an unprecedented combination of reliability, safety, durability, throttlability, and restart capabilities in a high-energy, cryogenic engine. The testbed selected for the deep throttling demonstration phases of this program was a minimally modified RL10 engine, allowing for maximum current production engine commonality and extensibility with minimum program cost. Two series of demonstrator engine tests, the first in April-May 2006 and the second in March-April 2007, have demonstrated in excess of 10:1 throttling of the hydrogen/oxygen expander cycle engine. Both test series have explored a combustion instability ("chug") environment at low throttled power levels. These tests have provided an early demonstration of an enabling cryogenic propulsion concept with invaluable system-level technology data acquisition toward design and development risk mitigation for future CECE Demonstrator engine tests.

Nuclear fuel performance: Trends, remedies and challenges

NASA Astrophysics Data System (ADS)

Rusch, C. A.

2008-12-01

It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

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

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

Influence of Localized Plasticity on IASCC Sensitivity of Austenitic Stainless Steels under PWR Primary Water

NASA Astrophysics Data System (ADS)

Cissé, Sarata; Tanguy, Benoit; Laffont, Lydia; Lafont, Marie-Christine; Guerre, Catherine; Andrieu, Eric

The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ɛp/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans <111>, that swept defects or γ' precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water

NASA Astrophysics Data System (ADS)

Fournier, L.; Savoie, M.; Delafosse, D.

2007-06-01

The low cycle fatigue (LCF) behaviour of precipitation-strengthened A-286 austenitic stainless steel was first investigated at room temperature under 0.2% plastic strain control. LCF led to hardening for the first 20 cycles and then to significant softening. LCF-induced dislocation microstructure was characterized using both bright and dark-field imaging techniques in transmission electron microscopy. Cycling softening was correlated with the formation of precipitate-free localized deformation bands. The effect of these precipitate-free localized deformation bands on A-286 stress corrosion cracking (SCC) behaviour in PWR primary water was then examined by means of constant extension rate tensile (CERT) tests at 320 °C and 360 °C. Comparative CERT tests were performed on companion specimens with similar yield stress but pre-fatigued to a few cycles (4-8) or between 125 and 200 cycles. Specimens pre-fatigued to a few cycles with no precipitate-free localized deformation bands exhibited little susceptibility to intergranular SCC (IGSCC). In contrast, the presence of precipitate-free localized deformation bands formed by pre-fatigue to between 125 and 200 cycles strongly promoted IGSCC. The interest of the approach used in this study is to provide insight into the role of localized deformation in irradiation assisted stress corrosion cracking.

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

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

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

2010-08-11

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

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

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

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

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.

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

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

Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiationmore » after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.« less

M3FT-16OR0203052-Test Design for FeCrAl Alloy Tube Irradiation in HFIR

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

Terrani, Kurt A.; Petrie, Christian M.

2016-05-01

This calculation summarizes thermal analyses of a flexible rabbit design for irradiating a variety of pressurized water reactor (PWR) cladding materials (stainless steel, iron-chromium aluminum [FeCrAl], Zircaloy, and Inconel) with variable dimensions at a temperature of 350 °C in the flux trap of the High Flux Isotope Reactor (HFIR). The design can accommodate standard cladding for outer diameters (ODs) of approximately 9.50 mm with thickness ranging from 0.30 mm to 0.70 mm. The length is generally between 10 and 50 mm. The specimens contain moly inserts with a variable OD that provides the heat flux necessary to achieve the designmore » temperature with such a small fixed gas gap. The primary outer containment is an Al-6061 housing with a slightly enlarged inner diameter (ID) of 9.60 mm. The specimen temperature is controlled by determining a helium/argon gas mixture specific to the as-built specimen and housing. Variables that affect the required gas mixture are the cladding material (thermal expansion, density, heat generation rate), cladding OD, housing ID, and cladding ID. This calculation documents the analyses performed to determine required gas mixtures for a variety of scenarios.« less

Microstructural evaluation of cumulative fatigue damage in a plant component sample

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

Fukuoka, C.; Nakagawa, Y.G.; Yoshida, K.

1996-12-31

Fatigue damage accumulated in a real plant was evaluated in terms of microstructural conditioning. Microstructural damage induced in laboratory by cyclic deformation near and below the fatigue limit was also examined. A Transmission Electron Microscopy (TEM) technique called the Selected Area Diffraction (SAD) method was employed in this study. In earlier studies, it was found that the SAD value indicating a magnitude of crystallographic misorientation in the substructure (dislocation cells) was increasing with the increase of fatigue damage accumulation. Small samples removed from PWR feed water nozzle welds were examined by the SAD. It was found that the damage statemore » measured by the SAD well agreed with the morphological evidence. Cyclic stresses near or below the fatigue limit were applied to samples taken from a SA508 steel plate at various stresses. The SAD value increased even below the fatigue limit, but there was no sign of microstructural conditioning below the stresses of 50% of the fatigue limit. These results suggested that at stresses below the current design curve (below half the fatigue limit) no microstructural conditioning proceeded. It was concluded that the microstructural method was effective to evaluate damage accumulation in real plant components, and also that the current design curve was adequate in terms of microstructural conditioning state.« less

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

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

David Andrs; Ray Berry; Derek Gaston

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

Westinghouse Small Modular Reactor balance of plant and supporting systems design

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

Memmott, M. J.; Stansbury, C.; Taylor, C.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operationmore » of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)« less

Conclusions and Recommendations Regarding the Deep Sea Hybrid Power Systems Initial Study

DTIC Science & Technology

2010-06-01

proton-exchange membrane fuel cells ( PEMFC ) powered with hydrogen and oxygen, similar to that used on proven subsurface vessels; (2) fuel-cells...AND STORAGE OPTIONS CONSIDERED FOR INITIAL STUDY NO. NOMENCLATURE DESCRIPTION 1 PWR Nuclear Reactor + Battery 2 FC1 PEMFC + Line for surface O2...Wellhead Gas + Reformer + Battery 3 FC2 PEMFC + Stored O2 + Wellhead Gas + Reformer + Battery 4 SV1 PEMFC + Submersible Vehicle for O2 Transport

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

DOE PAGES

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

2014-09-29

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

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

NASA Astrophysics Data System (ADS)

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

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

Distribution of the messenger RNA for the small conductance calcium-activated potassium channel SK3 in the adult rat brain and correlation with immunoreactivity.

PubMed

Tacconi, S; Carletti, R; Bunnemann, B; Plumpton, C; Merlo Pich, E; Terstappen, G C

2001-01-01

Small conductance calcium-activated potassium channels are voltage independent potassium channels which modulate the firing patterns of neurons by activating the slow component of the afterhyperpolarization. The genes encoding a family of small conductance calcium-activated potassium channels have been cloned and up to now three known members have been described and named small conductance calcium-activated potassium channel type 1, small conductance calcium-activated potassium channel type 2 and small conductance calcium-activated potassium channel type 3; the distribution of their messenger RNA in the rat CNS has already been performed but only in a limited detail. The present study represents the first detailed analysis of small conductance calcium-activated potassium channel type 3 mRNA distribution in the adult rat brain and resulted in a strong to moderate expression of signal in medial habenular nucleus, substantia nigra compact part, suprachiasmatic nucleus, ventral tegmental area, lateral septum, dorsal raphe and locus coeruleus. Immunohistological experiments were also performed and confirmed the presence of small conductance calcium-activated potassium channel type 3 protein in medial habenular nucleus, locus coeruleus and dorsal raphe. Given the importance of dorsal raphe, locus coeruleus and substantia nigra/ventral tegmental area for serotonergic, noradrenergic and dopaminergic transmission respectively, our results pose the morphological basis for further studies on the action of small conductance calcium-activated potassium channel type 3 in serotonergic, noradrenergic and dopaminergic transmission.

Effect of Different Types of Small-Group Activities on Students' Conversations

ERIC Educational Resources Information Center

Young, Krista K.; Talanquer, Vicente

2013-01-01

Teaching reform efforts in chemistry education often involve engaging students in small-group activities of different types. This study focused on the analysis of how activity type affected the nature of group conversations. In particular, we analyzed the small-group conversations of students enrolled in a chemistry course for nonscience majors.…

Effect of lifestyle interventions of pregnant women on their dietary habits, lifestyle behaviors, and weight gain: a randomized controlled trial.

PubMed

Aşcı, Özlem; Rathfisch, Gülay

2016-02-24

Although it is known that lifestyle behaviors of pregnant women are closely related to maternal and fetal health, number of data concerning efficacy of intervention on lifestyle during pregnancy is limited. The purpose of this study is to determine the effect of lifestyle interventions on improving dietary habits and lifestyle behaviors, ensuring gestational weight gain (GWG) within recommended levels and limiting postpartum weight retention (PWR). The study was conducted as a randomized controlled trial in a family health center located in Istanbul, Turkey, between June 2011 and July 2012. The primary outcomes were GWG, and the proportion of pregnant women whose GWG was within the Institute of Medicine (IOM) guidelines. One hundred two pregnant women with gestation ≤12 weeks, age ≥18 years, gravidity ≤2, and who did not intend to lose weight in prepregnancy period were randomly included in this study as intervention (n = 51) and control (n = 51) groups. The study was completed with 45 women for each group. The control group received routine antenatal care. The intervention group was received an individualized lifestyle intervention focusing on healthy lifestyle, diet, exercise, and weight monitoring as four sessions at 12-15, 16-18, 20-24, and 37 weeks gestation. Lifestyle behaviors were evaluated with Health-Promoting Lifestyle Profile-II. Dietary habits were assessed by 3-day dietary recalls, and weight was followed from pregnancy until 6 weeks postpartum. The lifestyle interventions had a significant effect on improving lifestyle behaviors, protein intake, percentage of energy from protein, calcium, magnesium, iron, zinc, and vegetable intakes when adjusted for confounders (p < 0.05). The proportion of women who were within the IOM recommendations was higher in the intervention group (51.1 %) than in the control group (28.9 %) The odds ratio for GWG within IOM was statistically significant between the groups (OR = 0.59, 95 % CI, 0.45-0.72). There were no difference between groups in terms of the other dietary intakes, total GWG, and PWR (p > 0.05). Lifestyle intervention improves the lifestyle behaviors during pregnancy and increases the appropriate GWG for prepregnancy body mass index (BMI), but it has a limited effect in terms of improving dietary habits and has no effect on PWR.

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

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

Liu, Maolong; Ryals, Matthew; Ali, Amir

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentallymore » investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.« less

On the Geological History of Venus

NASA Astrophysics Data System (ADS)

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

2008-09-01

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

A Comparative Study of Various Electric Propulsion Systems and their Impact on a Nominal Ship Design

DTIC Science & Technology

1987-06-01

Permanent Magnet Motors ," Advanced Mechanical Technology, Inc., 1983. 20. Marshall, McMurray, Richter, Webster, and...December 1977. 21. Triezenbarg, Greene, Hannan, and Dvorsky, "Study of Permanent Magnet Motors for Naval Propulsion," Westinghcuse Research Report 80=9B2...34 Paper 71 CP 155-PWR, IEEE Winter Power Meeting, New York, February 1971. 34. Ireland, James R., Ceramic Permanent - Magnet Motors , McGraw-Hill, New York, 1968. 206

Automotive Test Rig Final Design Report. Volume 2. Control System.

DTIC Science & Technology

1986-01-01

Pressure Switch Status P27 Low Brake Release Pressure Switch Status P26 Low Brake...Supply Pressure Switch Status P25 Low Port Charge Pump Pressure Switch Status P24 Low Starboard Charge Pump Pressure Switch Status P23 Hydraulic Filter By...Sensed Switch Status P31 Low Scavenge Pump Pressure Switch Status P30 P37 Signal Return for Computer J21 Not Used J22 P A +24 B Pwr Rtn C Ground C

Nuclear Fuel Depletion Analysis Using Matlab Software

NASA Astrophysics Data System (ADS)

Faghihi, F.; Nematollahi, M. R.

Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.

A NIST Kinetic Data Base for PAH Reaction and Soot Particle Inception During Combusion

DTIC Science & Technology

2007-12-01

in Computational Fluid Dynamics (CFD) codes hat have lead to the capability of describing complex reactive flow problems and thus simulating... parameters . However in the absence of data estimates must be made. Since the chemistry of combustion is extremely complex and for proper description...118:381-389 9. Babushok, V. and Tsang, W., J. Prop. and Pwr . 20 (2004) 403-414. 10. . Fournet, R., Warth, V., Glaude, P.A., Battin-Leclerc, F

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

Heat transfer during condensation of steam from steam-gas mixtures in the passive safety systems of nuclear power plants

NASA Astrophysics Data System (ADS)

Portnova, N. M.; Smirnov, Yu B.

2017-11-01

A theoretical model for calculation of heat transfer during condensation of multicomponent vapor-gas mixtures on vertical surfaces, based on film theory and heat and mass transfer analogy is proposed. Calculations were performed for the conditions implemented in experimental studies of heat transfer during condensation of steam-gas mixtures in the passive safety systems of PWR-type reactors of different designs. Calculated values of heat transfer coefficients for condensation of steam-air, steam-air-helium and steam-air-hydrogen mixtures at pressures of 0.2 to 0.6 MPa and of steam-nitrogen mixture at the pressures of 0.4 to 2.6 MPa were obtained. The composition of mixtures and vapor-to-surface temperature difference were varied within wide limits. Tube length ranged from 0.65 to 9.79m. The condensation of all steam-gas mixtures took place in a laminar-wave flow mode of condensate film and turbulent free convection in the diffusion boundary layer. The heat transfer coefficients obtained by calculation using the proposed model are in good agreement with the considered experimental data for both the binary and ternary mixtures.

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

PubMed

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

2010-05-01

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

Costs, equity, efficiency and feasibility of identifying the poor in Ghana's National Health Insurance Scheme: empirical analysis of various strategies.

PubMed

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

2012-01-01

To analyse the costs and evaluate the equity, efficiency and feasibility of four strategies to identify poor households for premium exemptions in Ghana's National Health Insurance Scheme (NHIS): means testing (MT), proxy means testing (PMT), participatory wealth ranking (PWR) and geographic targeting (GT) in urban, rural and semi-urban settings in Ghana. We conducted the study in 145-147 households per setting with MT as our gold standard strategy. We estimated total costs that included costs of household surveys and cost of premiums paid to the poor, efficiency (cost per poor person identified), equity (number of true poor excluded) and the administrative feasibility of implementation. The cost of exempting one poor individual ranged from US$15.87 to US$95.44; exclusion of the poor ranged between 0% and 73%. MT was most efficient and equitable in rural and urban settings with low-poverty incidence; GT was efficient and equitable in the semi-urban setting with high-poverty incidence. PMT and PWR were less equitable and inefficient although feasible in some settings. We recommend MT as optimal strategy in low-poverty urban and rural settings and GT as optimal strategy in high-poverty semi-urban setting. The study is relevant to other social and developmental programmes that require identification and exemptions of the poor in low-income countries. © 2011 Blackwell Publishing Ltd.

Korean standard nuclear plant ex-vessel neutron dosimetry program Ulchin 4

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

Duo, J.I.; Chen, J.; Kulesza, J.A.

2011-07-01

A comprehensive ex-vessel neutron dosimetry (EVND) surveillance program has been deployed in 16 pressurized water reactors (PWR) in South Korea and EVND dosimetry sets have already been installed and analyzed in Westinghouse reactor designs. In this paper, the unique features of the design, training, and installation in the Korean standard nuclear plant (KSNP) Ulchin Unit 4 are presented. Ulchin Unit 4 Cycle 9 represents the first dosimetry analyzed from the EVND design deployed in KSNP plants: Yonggwang Units 3 through 6 and Ulchin Units 3 through 6. KSNP's cavity configuration precludes a conventional installation from the cavity floor. The solution,more » requiring the installation crew to access the cavity at an elevation of the active core, places a premium on rapid installation due to high area dose rates. Numerous geometrical features warranted the use of a detailed design in true 3D mechanical design software to control interferences. A full-size training mockup maximized the crew ability to correctly install the instrument in minimum time. The analysis of the first dosimetry set shows good agreements between measurement and calculation within the associated uncertainties. A complete EVND system has been successfully designed, installed, and analyzed for a KNSP plant. Current and future EVND analyses will continue supporting the successful operation of PWR units in South Korea. (authors)« less

The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review

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

Rebak, R B; Hua, F H

2004-07-12

Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attemptmore » is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.« less

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

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.

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

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

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

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

Youinou, Gilles J.

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE PAGES

Youinou, Gilles J.

2017-05-04

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

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

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

Matthew Ellis; Derek Gaston; Benoit Forget

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

Constraints on silicates formation in the Si-Al-Fe system: Application to hard deposits in steam generators of PWR nuclear reactors

NASA Astrophysics Data System (ADS)

Berger, Gilles; Million-Picallion, Lisa; Lefevre, Grégory; Delaunay, Sophie

2015-04-01

Introduction: The hydrothermal crystallization of silicates phases in the Si-Al-Fe system may lead to industrial constraints that can be encountered in the nuclear industry in at least two contexts: the geological repository for nuclear wastes and the formation of hard sludges in the steam generator of the PWR nuclear plants. In the first situation, the chemical reactions between the Fe-canister and the surrounding clays have been extensively studied in laboratory [1-7] and pilot experiments [8]. These studies demonstrated that the high reactivity of metallic iron leads to the formation of Fe-silicates, berthierine like, in a wide range of temperature. By contrast, the formation of deposits in the steam generators of PWR plants, called hard sludges, is a newer and less studied issue which can affect the reactor performance. Experiments: We present here a preliminary set of experiments reproducing the formation of hard sludges under conditions representative of the steam generator of PWR power plant: 275°C, diluted solutions maintained at low potential by hydrazine addition and at alkaline pH by low concentrations of amines and ammoniac. Magnetite, a corrosion by-product of the secondary circuit, is the source of iron while aqueous Si and Al, the major impurities in this system, are supplied either as trace elements in the circulating solution or by addition of amorphous silica and alumina when considering confined zones. The fluid chemistry is monitored by sampling aliquots of the solution. Eh and pH are continuously measured by hydrothermal Cormet© electrodes implanted in a titanium hydrothermal reactor. The transformation, or not, of the solid fraction was examined post-mortem. These experiments evidenced the role of Al colloids as precursor of cements composed of kaolinite and boehmite, and the passivation of amorphous silica (becoming unreactive) likely by sorption of aqueous iron. But no Fe-bearing was formed by contrast to many published studies on the Fe-clay interactions in the nuclear waste storage, and by contrast with basic thermodynamic predictions. Conclusion: The Fe-clays and steam generators contexts imply relatively close aqueous environments: hydrothermal, reduced, diluted, neutral to slightly alkaline. The main difference is the status of iron: ferric/ferrous (magnetite) in the steam generators, metallic in the Fe-clay experiments. The concentration of aqueous iron when supplied by magnetite is low and does not allow its incorporation in secondary phases. By contrast, aqueous ferrous iron released by the corrosion of steel is not limited by the source, rather by the sink, and produces Fe-rich silicates. This example illustrates the discrepancy between complex mineral reactions and oversimplified predictions when sorption/passivation and nucleation/growth constraints are ignored. Reference: [1] Lanson et al. (2012) Amer. Min. 97, 864-871. [2] Lantenois et al. (2005) Clays & Clay Min. 53, 597-612. [3] Mosser-Ruck et al. (2010) Clays & Clay Min. 58, 280-291. [4] Perronnet et al. (2008) App. Clay Sci. 38, 187-202. [5] Osacky et al. (2010) App. Clay Sci. 50, 237-244. [6] Guillaume et al. (2003) Clay Min. 38, 281-302. [7] Rivard et al. (2013) Amer. Mineral. 98, 163-180. [8] Svensson and Hansen (2013) Clays & Clay Min. 61, 566-579.

Concomitant loss of SMARCA2 and SMARCA4 expression in small cell carcinoma of the ovary, hypercalcemic type.

PubMed

Jelinic, Petar; Schlappe, Brooke A; Conlon, Niamh; Tseng, Jill; Olvera, Narciso; Dao, Fanny; Mueller, Jennifer J; Hussein, Yaser; Soslow, Robert A; Levine, Douglas A

2016-01-01

Small cell carcinoma of the ovary, hypercalcemic type is an aggressive tumor generally affecting young women with limited treatment options. Mutations in SMARCA4, a catalytic subunit of the SWI/SNF chromatin remodeling complex, have recently been identified in nearly all small cell carcinoma of the ovary, hypercalcemic type cases and represent a signature molecular feature for this disease. Additional biological dependencies associated with small cell carcinoma of the ovary, hypercalcemic type have not been identified. SMARCA2, another catalytic subunit of the SWI/SNF complex mutually exclusive with SMARCA4, is thought to be post-translationally silenced in various cancer types. We analyzed 10 archival small cell carcinoma of the ovary, hypercalcemic type cases for SMARCA2 protein expression by immunohistochemistry and found that SMARCA2 expression was lost in all but one case. None of the 50 other tumors that primarily or secondarily involved the ovary demonstrated concomitant loss of SMARCA2 and SMARCA4. Deep sequencing revealed that this loss of SMARCA2 expression is not the result of mutational inactivation. In addition, we established a small cell carcinoma of the ovary, hypercalcemic type patient-derived xenograft and confirmed the loss of SMARCA2 in this in vitro model. This patient-derived xenograft model, established from a recurrent tumor, also had unexpected mutational features for this disease, including functional mutations in TP53 and POLE. Taken together, our data suggest that concomitant loss of SMARCA2 and SMARCA4 is another hallmark of small cell carcinoma of the ovary, hypercalcemic type-a finding that offers new opportunities for therapeutic interventions.

Recent advances in developing small molecules targeting RNA.

PubMed

Guan, Lirui; Disney, Matthew D

2012-01-20

RNAs are underexploited targets for small molecule drugs or chemical probes of function. This may be due, in part, to a fundamental lack of understanding of the types of small molecules that bind RNA specifically and the types of RNA motifs that specifically bind small molecules. In this review, we describe recent advances in the development and design of small molecules that bind to RNA and modulate function that aim to fill this void.

Land use determinants of small mammal abundance and distribution in a plague endemic area of Lushoto District, Tanzania.

PubMed

Hieronimo, Proches; Kimaro, Didas N; Kihupi, Nganga I; Gulinck, Hubert; Mulungu, Loth S; Msanya, Balthazar M; Leirs, Herwig; Deckers, Jozef A

2014-07-01

Small mammals are considered to be involved in the transmission cycle of bubonic plague, still occurring in different parts of the world, including the Lushoto District in Tanzania. The objective of this study was to determine the relationship between land use types and practices and small mammal abundance and distribution. A field survey was used to collect data in three landscapes differing in plague incidences. Data collection was done both in the wet season (April-June 2012) and dry season (August-October 2012). Analysis of variance and Boosted Regression Trees (BRT) modelling technique were used to establish the relationship between land use and small mammal abundance and distribution. Significant variations (p ≤ 0.05) of small mammal abundance among land use types were identified. Plantation forest with farming, natural forest and fallow had higher populations of small mammals than the other aggregated land use types. The influence of individual land use types on small mammal abundance level showed that, in both dry and wet seasons, miraba and fallow tended to favour small mammals' habitation whereas land tillage practices had the opposite effect. In addition, during the wet season crop types such as potato and maize appeared to positively influence the distribution and abundance of small mammals which was attributed to both shelter and food availability. Based on the findings from this study it is recommended that future efforts to predict and map spatial and temporal human plague infection risk at fine scale should consider the role played by land use and associated human activities on small mammal abundance and distribution.

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

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

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

Motley, F.

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

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

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

Not Available

1979-07-01

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

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

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

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

1983-01-01

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

Crack growth testing on Cold Worked Alloy 690 in Primary Water Environment

NASA Astrophysics Data System (ADS)

Tice, David R.; Medway, Stuart L.; Platts, Norman; Stairmand, John W.

While plant experience so far has shown excellent resistance of Alloy 690 to stress corrosion cracking in PWR primary water environments, laboratory tests have reported that susceptibility may be enhanced substantially by non-uniform cold working, particularly when the plane of crack growth is in the plane of rolling or forging. The Alloy 690 program aims to further the understanding of the mechanisms behind this susceptibility and the heat-to-heat variability reported for different materials.

Advanced Information Systems Design: Technical Basis and Human Factors Review Guidance

DTIC Science & Technology

2000-03-01

D ., Wise, J ., and Hanes, L., "An Evaluation of Nuclear Power Plant Safety Parameter Display Systems," Proceedings of the Human Factors Society 25th...Reactor (PWR) (Source: Reprinted with permission from Woods, D ., Wise, J ., and Hanes, L., "An Evaluation of Nuclear Power Plant Safety Parameter...Dials display rpCJni?3 (b) Fluid-Tanks display B (c) Seesaw display I 72 CF \\^- J B ’ V ’II ’ ( d ) Mimic display B E * • \\ ^r 7

Hot zero power reactor calculations using the Insilico code

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; ...

2016-03-18

In this paper we describe the reactor physics simulation capabilities of the insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that the insilico SP N solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various PWR problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

33 CFR 208.11 - Regulations for use of storage allocated for flood control or navigation and/or project operation...

Code of Federal Regulations, 2011 CFR

2011-07-01

... VA Pwr. Glen Elder Dam & Waconda Lk KS Mitchel Solomon R FIM 722.3204.8 1488.31455.6 1455.61428.0... 820 PL 78-534 USBR. Kirwin Dam & Res KS Phillips N Fork Solomon R F ICR 215.1 89.6 1757.3 1729.2 1729... Webster Dam & Res KS Rocks S Fork Solomon R F IRC 183.4 72.1 1923.7 1892.5 1892.5 1860.0 8480 3772 3772...

33 CFR 208.11 - Regulations for use of storage allocated for flood control or navigation and/or project operation...

Code of Federal Regulations, 2010 CFR

2010-07-01

... VA Pwr. Glen Elder Dam & Waconda Lk KS Mitchel Solomon R FIM 722.3204.8 1488.31455.6 1455.61428.0... 820 PL 78-534 USBR. Kirwin Dam & Res KS Phillips N Fork Solomon R F ICR 215.1 89.6 1757.3 1729.2 1729... Webster Dam & Res KS Rocks S Fork Solomon R F IRC 183.4 72.1 1923.7 1892.5 1892.5 1860.0 8480 3772 3772...

Advanced techniques for repair of irradiated PWR fuel assemblies

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

Knaab, H.; Westphal, M.

Kraftwerk union has recently designed and built a portable repair unit for use in nuclear power plants for repair of defective fuel assemblies where space limitations do not allow permanent installation of repair equipment. This new equipment is designed to be easily disassembled and decontaminated. The main component of the equipment is the fuel assembly reconstitution unit (FARU) which is placed on the floor of the spent fuel pool. The use of the FARU is described in the paper.

Survey and Analysis of Environmental Requirements for Shipboard Electronic Equipment Applications. Appendix A. Volume 2.

DTIC Science & Technology

1991-07-31

INTELLIGENT SCSI DMV-719 MAS MIL CONTROLLER DY-4 SYSTEMS BYTE-WIDE MEMORY CARD DMV-536 MEM MIL DY-4 SYSTEMS POWER SUPPLY UNIT DMV-870 PWR MIL P age No. 5 06/10...FORCE COMPUTERS PROCESSOR CPU-386 SERIES SBC COM FORCE COMPUTERS ADVANCED SYSTEM CONTROL ASCU -1/2 SBC COM UNITI FORCE COMPUTERS GRAPHICS CONTROLLER AGC...RECORD VENDOR: JANZ COMPUTER AG DIVISION: VENDOR ADDRESS: Im Doerener Feld 3 D-4790 Paderborn Germany MARKETING: Johannes Kunz TECHNICAL: Arnulf

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.

JPRS Report Science & Technology Japan

DTIC Science & Technology

1989-03-02

Oxychlorides MOCln_2 (Organic Metal Salts) Alkoxides M(OR)n Acetylacetonate M(C5H702)n Acetates M(C2H302)n Oxalates M(C204)n/2 2.2 Hydrolysis and Gel...more deeply understanding hydrothermal dynamics during not only a major rupture LOCA but also a minor rupture LOCA and clarifying the combination of... hydrothermal dynamics of the coolant from the beginning of LOCA to its end, using a scale model of PWR (pressurized water reactor). Under the ROSA-III Plan

Precursor Evolution and Stress Corrosion Cracking Initiation of Cold-Worked Alloy 690 in Simulated Pressurized Water Reactor Primary Water

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

Zhai, Ziqing; Toloczko, Mychailo; Kruska, Karen

Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 (UNS N06690) materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for either the 21% or 31%CW CLT specimens loaded at their yield stress after ~9,220 hours, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showedmore » DCPD-indicated crack initiation after 10,400 hours of exposure at constant stress intensity, which was resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and discusses their effects on crack initiation in CW alloy 690.« less

Identification and quantitation of morphological cell types in electrophoretically separated human embryonic kidney cell cultures

NASA Technical Reports Server (NTRS)

Williams, K. B.; Kunze, M. E.; Todd, P. W.

1985-01-01

Four major cell types were identified by phase microscopy in early passage human embryonic kidney cell cultures. They are small and large epithelioid, domed, and fenestrated cells. Fibroblasts are also present in some explants. The percent of each cell type changes with passage number as any given culture grows. As a general rule, the fraction of small epithelioid cells increases, while the fraction of fenestrated cells, always small, decreases further. When fibroblasts are present, they always increase in percentage of the total cell population. Electrophoretic separation of early passage cells showed that the domed cells have the highest electrophoretic mobility, fibroblasts have an intermediate high mobility, small epithelioid cells have a low mobility, broadly distributed, and fenestrated cells have the lowest mobility. All cell types were broadly distributed among electrophoretic subfractions, which were never pure but only enriched with respect to a given cell type.

Small Business Participation in Air Force Procurement: Participation Trends and the Effect of Acquisition Reform Initiatives on Air Force Procurement via Selected Contract Types in Selected Product Categories

DTIC Science & Technology

2004-06-01

directed the SBA to provide several types of assistance to small businesses , including technical and management assistance, loans, and assistance in...Directed the SBA to provide small businesses with technical and management assistance - Authorized the SBA to enter into contracts with Federal...command and center directors of small business , small business specialists, data managers , and Air Force Institute of Technology student researchers

Concomitant loss of SMARCA2 and SMARCA4 expression in small cell carcinoma of the ovary, hypercalcemic type

PubMed Central

Jelinic, Petar; Schlappe, Brooke A; Conlon, Niamh; Tseng, Jill; Olvera, Narciso; Dao, Fanny; Mueller, Jennifer J; Hussein, Yaser; Soslow, Robert A; Levine, Douglas A

2016-01-01

Small cell carcinoma of the ovary, hypercalcemic type is an aggressive tumor generally affecting young women with limited treatment options. Mutations in SMARCA4, a catalytic subunit of the SWI/SNF chromatin remodeling complex, have recently been identified in nearly all small cell carcinoma of the ovary, hypercalcemic type cases and represent a signature molecular feature for this disease. Additional biological dependencies associated with small cell carcinoma of the ovary, hypercalcemic type have not been identified. SMARCA2, another catalytic subunit of the SWI/SNF complex mutually exclusive with SMARCA4, is thought to be post-translationally silenced in various cancer types. We analyzed 10 archival small cell carcinoma of the ovary, hypercalcemic type cases for SMARCA2 protein expression by immunohistochemistry and found that SMARCA2 expression was lost in all but one case. None of the 50 other tumors that primarily or secondarily involved the ovary demonstrated concomitant loss of SMARCA2 and SMARCA4. Deep sequencing revealed that this loss of SMARCA2 expression is not the result of mutational inactivation. In addition, we established a small cell carcinoma of the ovary, hypercalcemic type patient-derived xenograft and confirmed the loss of SMARCA2 in this in vitro model. This patient-derived xenograft model, established from a recurrent tumor, also had unexpected mutational features for this disease, including functional mutations in TP53 and POLE. Taken together, our data suggest that concomitant loss of SMARCA2 and SMARCA4 is another hallmark of small cell carcinoma of the ovary, hypercalcemic type—a finding that offers new opportunities for therapeutic interventions. PMID:26564006

Regulation of Small Mitochondrial DNA Replicative Advantage by Ribonucleotide Reductase in Saccharomyces cerevisiae

PubMed Central

Bradshaw, Elliot; Yoshida, Minoru; Ling, Feng

2017-01-01

Small mitochondrial genomes can behave as selfish elements by displacing wild-type genomes regardless of their detriment to the host organism. In the budding yeast Saccharomyces cerevisiae, small hypersuppressive mtDNA transiently coexist with wild-type in a state of heteroplasmy, wherein the replicative advantage of the small mtDNA outcompetes wild-type and produces offspring without respiratory capacity in >95% of colonies. The cytosolic enzyme ribonucleotide reductase (RNR) catalyzes the rate-limiting step in dNTP synthesis and its inhibition has been correlated with increased petite colony formation, reflecting loss of respiratory function. Here, we used heteroplasmic diploids containing wild-type (rho+) and suppressive (rho−) or hypersuppressive (HS rho−) mitochondrial genomes to explore the effects of RNR activity on mtDNA heteroplasmy in offspring. We found that the proportion of rho+ offspring was significantly increased by RNR overexpression or deletion of its inhibitor, SML1, while reducing RNR activity via SML1 overexpression produced the opposite effects. In addition, using Ex Taq and KOD Dash polymerases, we observed a replicative advantage for small over large template DNA in vitro, but only at low dNTP concentrations. These results suggest that dNTP insufficiency contributes to the replicative advantage of small mtDNA over wild-type and cytosolic dNTP synthesis by RNR is an important regulator of heteroplasmy involving small mtDNA molecules in yeast. PMID:28717049

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

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

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

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

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

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

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

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

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

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

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

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

Development and Testing of Neutron Cross Section Covariance Data for SCALE 6.2

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

Marshall, William BJ J; Williams, Mark L; Wiarda, Dorothea

2015-01-01

Neutron cross-section covariance data are essential for many sensitivity/uncertainty and uncertainty quantification assessments performed both within the TSUNAMI suite and more broadly throughout the SCALE code system. The release of ENDF/B-VII.1 included a more complete set of neutron cross-section covariance data: these data form the basis for a new cross-section covariance library to be released in SCALE 6.2. A range of testing is conducted to investigate the properties of these covariance data and ensure that the data are reasonable. These tests include examination of the uncertainty in critical experiment benchmark model k eff values due to nuclear data uncertainties, asmore » well as similarity assessments of irradiated pressurized water reactor (PWR) and boiling water reactor (BWR) fuel with suites of critical experiments. The contents of the new covariance library, the testing performed, and the behavior of the new covariance data are described in this paper. The neutron cross-section covariances can be combined with a sensitivity data file generated using the TSUNAMI suite of codes within SCALE to determine the uncertainty in system k eff caused by nuclear data uncertainties. The Verified, Archived Library of Inputs and Data (VALID) maintained at Oak Ridge National Laboratory (ORNL) contains over 400 critical experiment benchmark models, and sensitivity data are generated for each of these models. The nuclear data uncertainty in k eff is generated for each experiment, and the resulting uncertainties are tabulated and compared to the differences in measured and calculated results. The magnitude of the uncertainty for categories of nuclides (such as actinides, fission products, and structural materials) is calculated for irradiated PWR and BWR fuel to quantify the effect of covariance library changes between the SCALE 6.1 and 6.2 libraries. One of the primary applications of sensitivity/uncertainty methods within SCALE is the assessment of similarities between benchmark experiments and safety applications. This is described by a c k value for each experiment with each application. Several studies have analyzed typical c k values for a range of critical experiments compared with hypothetical irradiated fuel applications. The c k value is sensitive to the cross-section covariance data because the contribution of each nuclide is influenced by its uncertainty; large uncertainties indicate more likely bias sources and are thus given more weight. Changes in c k values resulting from different covariance data can be used to examine and assess underlying data changes. These comparisons are performed for PWR and BWR fuel in storage and transportation systems.« less

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

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

De Rosa, Felice

2006-07-01

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

Small bowel volvulus with jejunal diverticulum: Primary or secondary?

PubMed

Shen, Xiao-Fei; Guan, Wen-Xian; Cao, Ke; Wang, Hao; Du, Jun-Feng

2015-09-28

Small bowel volvulus, which is torsion of the small bowel and its mesentery, is a medical emergency, and is categorized as primary or secondary type. Primary type often occurs without any apparent intrinsic anatomical anomalies, while the secondary type is common clinically and could be caused by numerous factors including postoperative adhesions, intestinal diverticulum, and/or tumors. Here, we report a rare case of a 60-year-old man diagnosed with small bowel volvulus using multidetector computed tomography (MDCT) angiography. Further discovery by laparotomy showed one jejunal diverticulum, longer corresponding mesentery with a narrower insertion, and a lack of mesenteric fat. This case report includes several etiological factors of small bowel volvulus, and we discuss the possible cause of small bowel volvulus in this patient. We also highlight the importance of MDCT angiography in the diagnosis of volvulus and share our experience in treating this disease.

Small bowel volvulus with jejunal diverticulum: Primary or secondary?

PubMed Central

Shen, Xiao-Fei; Guan, Wen-Xian; Cao, Ke; Wang, Hao; Du, Jun-Feng

2015-01-01

Small bowel volvulus, which is torsion of the small bowel and its mesentery, is a medical emergency, and is categorized as primary or secondary type. Primary type often occurs without any apparent intrinsic anatomical anomalies, while the secondary type is common clinically and could be caused by numerous factors including postoperative adhesions, intestinal diverticulum, and/or tumors. Here, we report a rare case of a 60-year-old man diagnosed with small bowel volvulus using multidetector computed tomography (MDCT) angiography. Further discovery by laparotomy showed one jejunal diverticulum, longer corresponding mesentery with a narrower insertion, and a lack of mesenteric fat. This case report includes several etiological factors of small bowel volvulus, and we discuss the possible cause of small bowel volvulus in this patient. We also highlight the importance of MDCT angiography in the diagnosis of volvulus and share our experience in treating this disease. PMID:26420976

Assessment and Management of Aging in Phenix Nuclear Power Plant

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

Dumarcher, V.; Bourrier, J.L.; Chaucheprat, P.

2006-07-01

The combination of one or several processes of ruins can involve the materials failure of a nuclear power plant. These processes arise from the external agents action such as the pressure, the mechanical efforts, the heat flows and the radiations constitute the whole of the 'actions' of the surrounding medium. The prolongation and the repetition of these effects can involve a deterioration of the machine. In accordance with the decree of February 26, 1974, the PWR operator must be firstly, sure that the system is controlled according to the situations considered in the file of dimensioning and secondly, be ablemore » to know anytime the life of the equipment. The physical phenomena which cause the structures ruin are less complex in the PWR than in the SFR. In the SFR, the high temperatures imposed on components for long periods can involve a significant creep. In the course of time, this deformations accelerate the release of fatigue cracks. To consider the creep, the reactor lifespan is correlated at the numbers of thermals transients envisaged initially. To realize the management of aging in Phenix power plant, it is necessary to carry out an individualized monitoring of the structures and not only on the vessel. We must ensure the good state and/or the correct operation of the significant stations for safety which are the control of the reactivity, the movement of control rods, the primary sodium containment and the decay heat removal. For that, we monitor the main vessel, the conical skirt, the IHX and the Core Cover Plug. A profound knowledge of the thermal transients of the past is necessary to carry out an effective assessment. In order to guarantee that any harmful situation is well taken into the management of aging, we monitor permanently certain measurements (primary and secondary pump speed, hot and cold pool temperatures, IHX-main vessel and reactor roof temperatures). We present in the article the scientific method used in the Physics Section. A logical diagram specific to the type of situation and the structure allows to associate the harmful transient at a identical situation which has been happened in the past. During the last two cycles, the nuclear power plant has sustained 34 startup (20 during the 51. cycle and 14 during the 52. cycle). After two cycles of operation, there is approximately 70 to 80% of occurrences authorized for the whole of the structures. For the last 4 cycles, the number of transients to come will remain quite lower than the number dimensioned initially. (authors)« less

TREAT Neutronics Analysis and Design Support, Part I: Multi-SERTTA

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

Bess, John D.; Woolstenhulme, Nicolas E.; Hill, Connie M.

2016-08-01

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis ofmore » a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.74 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies.« less

Dissociation of retinoblastoma gene protein hyperphosphorylation and commitment to enter S phase.

PubMed Central

Ogryzko, V V; Hirai, T H; Shih, C E; Howard, B H

1994-01-01

Mitogenic activities of simian virus 40 large T and small t antigens were studied in serum-deprived human diploid fibroblasts. Wild-type large T and small t cooperated in stimulating DNA synthesis and in inducing hyperphosphorylation of the Rb gene product (pRb). In contrast, a T antigen mutant defective for pRb binding (Rb- T) possessed no detectable mitogenic activity alone and failed to complement small t in stimulating DNA synthesis. Surprisingly, Rb- T and small t cooperated as strongly as wild-type T and small t with respect to pRb hyperphosphorylation. As a consequence, in two closely related conditions (i.e., stimulation by small t plus wild-type T versus small t plus Rb- T), the fraction of pRb in hyperphosphorylated forms dissociated from the fraction of cells in the S phase. These results indicate that pRb hyperphosphorylation is not always tightly coupled with a commitment to initiate DNA replication. Images PMID:8189510

Etoposide Injection

MedlinePlus

... medications to treat a certain type of lung cancer (small cell lung cancer; SCLC). Etoposide is in a class of medications ... organs where eggs are formed), another type of lung cancer (non-small cell lung cancer; NSCLC), and Kaposi's ...

Retrofit designs for small bench-type blood cell counters.

PubMed

Ferris, C D

1991-01-01

This paper describes several retrofit designs to correct operational problems associated with small bench-type blood cell counters. Replacement electronic circuits as well as modifications to the vacuum systems are discussed.

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

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

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

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

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

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

Hendrix, C.E.

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

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

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

Szilard, Ronaldo Henriques

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

Integrated Systems Performance Assessment for the Evaluation of Space Nuclear Reactor Design Concepts (Phase 1: Demonstration of the Methodology).

DTIC Science & Technology

1992-11-01

Incorporated. Each design is characterized by a moderated core, a NaK pumped loop primary coolant system, and a potassium heat pipe radiator as the...1 1 10 1 RelHX 1 2 10 2 nRel HX 3 3 RelSS nRelSS Irr 4 3 7 8 9 io 2 + 2 + 2 + 2 nRel Pwr nRel NaK nRel RC nRel HX 1 1 11 1 RelSS 1 2 11 2 nRel SS 3 3

Use of artificial intelligence in severe accident diagnosis for PWRs

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

Wu, Zheng; Okrent, D.; Kastenberg, W.E.

1995-12-31

A combination approach of an expert system and neural networks is used to implement a prototype severe accident diagnostic system which would monitor the progression of the severe accident and provide necessary plant status information to assist the plant staff in accident management during the accident. The station blackout accident in a pressurized water reactor (PWR) is used as the study case. The current phase of research focus is on distinguishing different primary system failure modes and following the accident transient before and up to vessel breach.

Pool-site fuel inspection and examination techniques applied by the Kraftwerk Union Aktiengesellschaft Fuel Service. [PWR; BWR

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

Knaab, H.; Knecht, K.

The need for pool-site inspection and examination of fuel assemblies was recognized by Kraftwerk Union Aktiengesellschaft with the commissioning of the first nuclear power stations. A wet sipping method has demonstrated high reliability in detection of leaking fuel assemblies. The visual inspection system is a versatile tool. It can be supplemented by attaching devices for oxide thickness measurement or surface replication. Repair of leaking pressurized water reactor fuel assemblies has improved fuel utilization. Applied methods and typical results are described.

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

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

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

Conceptual Designing of a Reduced Moderation Pressurized Water Reactor by Use of MVP and MVP-BURN

NASA Astrophysics Data System (ADS)

Kugo, T.

A conceptual design of a seed-blanket assembly PWR core with a complicated geometry and a strong heterogeneity has been carried forward by use of the continuous-energy Monte Carlo method. Through parametric survey calculations by repeated use of MVP and a lattice burn-up calculation by MVP-BURN, a seed-blanket assembly configuration suitable for a concept of RMWR has been established, by evaluating precisely reactivity, a conversion ratio and a coolant void reactivity coefficient in a realistic computation time on a super computer.

Development of 3D pseudo pin-by-pin calculation methodology in ANC

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

Zhang, B.; Mayhue, L.; Huria, H.

2012-07-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. Themore » mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)« less

Nonlinear dynamic systems identification using recurrent interval type-2 TSK fuzzy neural network - A novel structure.

PubMed

El-Nagar, Ahmad M

2018-01-01

In this study, a novel structure of a recurrent interval type-2 Takagi-Sugeno-Kang (TSK) fuzzy neural network (FNN) is introduced for nonlinear dynamic and time-varying systems identification. It combines the type-2 fuzzy sets (T2FSs) and a recurrent FNN to avoid the data uncertainties. The fuzzy firing strengths in the proposed structure are returned to the network input as internal variables. The interval type-2 fuzzy sets (IT2FSs) is used to describe the antecedent part for each rule while the consequent part is a TSK-type, which is a linear function of the internal variables and the external inputs with interval weights. All the type-2 fuzzy rules for the proposed RIT2TSKFNN are learned on-line based on structure and parameter learning, which are performed using the type-2 fuzzy clustering. The antecedent and consequent parameters of the proposed RIT2TSKFNN are updated based on the Lyapunov function to achieve network stability. The obtained results indicate that our proposed network has a small root mean square error (RMSE) and a small integral of square error (ISE) with a small number of rules and a small computation time compared with other type-2 FNNs. Copyright © 2017 ISA. Published by Elsevier Ltd. All rights reserved.

Structural Maps of the V-17 Beta Regio Quadrangle, Venus

NASA Technical Reports Server (NTRS)

Basilevsky, A. t.; Head, James W.

2008-01-01

These represent slices of the geologic map into 7 time-stratigraphic levels whose descriptions are found in [3-6]. From older to younger they are: 1) Tessera material unit (t), 2) Densely fractured plains material unit (pdf), 3) Fractured and ridged plains material unit (pfr), 4) Tessera transitional terrain structural unit (tt), 5) Fracture belts structural unit (fb), 6) Shield plains (psh) and plains with wrinkle ridges (pwr) material units combined, and 7) Lobate (pl) and smooth (ps) plains material units combined and, approximately contemporaneous with them, the structural unit of rifted terrain (rt). Each slice shows the generalized pattern of structures typical of these units. Figures 1-7 show the seven maps and Figure 8 shows the combined map illustrating what is shown in the seven maps. To visualize the Beta Regio uplift outlines, the major structure of this area, we show the +0.5 km and +2.5 km contour lines, corresponding respectively to the base and the mid-height of the uplift. It is seen in Figures 1-2 and 4 the trends of t, pdf and tt occupy relatively small areas and their structures seen in these small windows appear rather variable and with almost no orientation heritage with time. Figure 3 shows that swarms of ridge belts trend mostly NW and go through the Beta structure with no alignment with it, suggesting that this structure did not yet exist at this time. Figure 5 shows that fracture belts align along the northern base of the Beta uplift suggesting onset of the formation of this structure. Figure 6 shows that wrinkle ridges do not show alignment with the Beta uplift suggesting that this already forming structure was not high enough to exert topographic stress in its vicinity. Figure 7 shows that the Beta uplift has Devana Chasma as an axial rift zone, suggesting a genetic link between the uplift and rifting. Figure 8 shows that structural trends in this area significantly changed with time.

Types and numbers of sensilla on antennae and maxillary palps of small and large houseflies, Musca domestica (Diptera, Muscidae).

PubMed

Smallegange, Renate C; Kelling, Frits J; Den Otter, Cornelis J

2008-12-01

Houseflies, Musca domestica, obtained from a high-larval-density culture were significantly (ca. 1.5 times) smaller than those from a low-larval-density culture. The same held true for their antennae and maxillary palps. Structure, number, and distribution of sensilla on antennae and palps of small and large flies were investigated using Scanning electron microscopy and Transmission electron microscopy. In each funiculus three pits were present, two (Type I) consisting of several compartments and one (Type II) of one compartment. Four types of olfactory sensilla were detected: trichoid sensilla on the funiculi, basiconic sensilla on funiculi and palps, grooved sensilla on funiculi and in pits Type I, and clavate sensilla on funiculi and in pits Type II. Type I pits also contained striated sensilla (presumably hygroreceptors). Mechanosensory bristles were present on scapes, pedicels, and palps. Noninnervated microtrichia were found on the palps and all antennal segments. The large houseflies possessed nearly twice as much sensilla as the small flies. So far, we did not observe differences in behavior between small and large flies. We assumed that small flies, being olfactory less equipped than large flies, may be able to compensate for this by, e.g., visual cues or by their olfactory sensilla being more sensitive than those of large flies. To be able to answer these questions careful studies have to be done on the behavioral responses of small and large flies to environmental stimuli. In addition, electrophysiological studies should be performed to reveal whether the responses of individual sensilla of flies reared under different conditions have been changed. 2008 Wiley-Liss, Inc.

Epidermal growth factor receptor gene mutation defines distinct subsets among small adenocarcinomas of the lung.

PubMed

Haneda, Hiroshi; Sasaki, Hidefumi; Shimizu, Shigeki; Endo, Katsuhiko; Suzuki, Eriko; Yukiue, Haruhiro; Kobayashi, Yoshihiro; Yano, Motoki; Fujii, Yoshitaka

2006-04-01

Epidermal growth factor receptor (EGFR) gene mutations are frequently detected in lung cancer, especially in adenocarcinoma, in females, and non-smoking patients. EGFR mutations are closely associated with clinical response to EGFR tyrosine kinase inhibitor. Bronchioloalveolar carcinoma (BAC) appearance is a good predictor of response to this agent. Noguchi et al. subdivided small peripheral adenocarcinoma of the lung into two groups. One group was characterized with tumor cell growth replacing the normal alveolar cells with varying degree of fibrosis (types A-C), and the other shows non-replacing and destructive growth (types D-F). Using probes for the 13 mutations which have been previously described, we have genotyped the EGFR gene status in surgically resected atypical adenomatous hyperplasias (AAH) and small peripheral adenocarcinomas up to 2 cm in diameter using TaqMan PCR assay. In 95 small-sized adenocarcinomas, the EGFR mutations were detected in 37 patients (38.9%), and no mutations were found in five AAHs. In small peripheral adenocarcinomas, EGFR mutations were found 47.1% of types A, B, or C adenocarcinomas; it was less frequent (16%) in Noguchi's types D, E or F adenocarcinomas. These results suggest that type D, F adenocarcinomas are not derived from the less malignant types A-C adenocarcinomas; rather, they have arisen de novo by distinct mechanisms. Although types A and B adenocarcinomas are almost 100% cured by surgery, some type C adenocarcinoma show lymph node metastasis and relapse. EGFR mutation analysis may help identify patients who will respond to treatment with tyrosine kinase inhibitors, e.g., gefitinib.

Solution Processed Organic Photovoltaic Cells Using D-A-D-A-D Type Small Molecular Donor Materials with Benzodithiophene and Diketopyrrolopyrrole Units.

PubMed

Park, Sangman; Nam, So Yeon; Suh, Dong Hack; Lee, Jaemin; Lee, Changjin; Yoon, Sung Cheol

2016-03-01

Organic photovoltaic Cells (OPVs) have been considered to be a next-generation energy source to overcome exhaustion of resources. Currently, OPVs are developed based on two types of donor material with polymer and small molecule. Polymeric donor materials have shown better power conversion efficiency (PCE) than small molecular donor materials, since it's easy to control the morphology of photoactive film. However, the difficulty in synthetic reproducibility and purification of polymeric donor were main drawback to overcome. And then, recently small molecule donor materials have been overcome bad morphology of OPVs film by using appropriate alkyl substituents and relatively long conjugation system. In this study, we designed and synthesized D-A-D-A-D type small molecular donor materials containing alternatively linked benzodithiophene (BDT) and diketopyrrolopyrrole (DPP) units. Also, we studied on the effect of photovoltaic performance of prepared small molecular D-A-D-A-D type donor with variation of thiophene links and with/without hexyl substituent. Our small molecular donors showed HOMO energy levels from -5.26 to -5.34 eV and optical bandgaps from 1.70 to 1.87 eV by CV (cyclic voltammetry) and UV/Vis spectroscopy, respectively. Finally, 3.4% of PCE can be obtained using a mixture of BDT(DPP)2-T2 and PCBM as an active layer with a Voc of 0.78 V, a Jsc of 9.72 mA/cm2, and a fill factor of 0.44 under 100 mW/cm2 AM 1.5G simulated light. We will discuss the performance of D-A-D-A-D type small molecular donor based OPVs with variation of both terminal substituents.

Do medical students generate sound arguments during small group discussions in problem-based learning?: an analysis of preclinical medical students' argumentation according to a framework of hypothetico-deductive reasoning.

PubMed

Ju, Hyunjung; Choi, Ikseon; Yoon, Bo Young

2017-06-01

Hypothetico-deductive reasoning (HDR) is an essential learning activity and a learning outcome in problem-based learning (PBL). It is important for medical students to engage in the HDR process through argumentation during their small group discussions in PBL. This study aimed to analyze the quality of preclinical medical students' argumentation according to each phase of HDR in PBL. Participants were 15 first-year preclinical students divided into two small groups. A set of three 2-hour discussion sessions from each of the two groups during a 1-week-long PBL unit on the cardiovascular system was audio-recorded. The arguments constructed by the students were analyzed using a coding scheme, which included four types of argumentation (Type 0: incomplete, Type 1: claim only, Type 2: claim with data, and Type 3: claim with data and warrant). The mean frequency of each type of argumentation according to each HDR phase across the two small groups was calculated. During small group discussions, Type 1 arguments were generated most often (frequency=120.5, 43%), whereas the least common were Type 3 arguments (frequency=24.5, 8.7%) among the four types of arguments. The results of this study revealed that the students predominantly made claims without proper justifications; they often omitted data for supporting their claims or did not provide warrants to connect the claims and data. The findings suggest instructional interventions to enhance the quality of medical students' arguments in PBL, including promoting students' comprehension of the structure of argumentation for HDR processes and questioning.

Formalizing Evidence Type Definitions for Drug-Drug Interaction Studies to Improve Evidence Base Curation.

PubMed

Utecht, Joseph; Brochhausen, Mathias; Judkins, John; Schneider, Jodi; Boyce, Richard D

2017-01-01

In this research we aim to demonstrate that an ontology-based system can categorize potential drug-drug interaction (PDDI) evidence items into complex types based on a small set of simple questions. Such a method could increase the transparency and reliability of PDDI evidence evaluation, while also reducing the variations in content and seriousness ratings present in PDDI knowledge bases. We extended the DIDEO ontology with 44 formal evidence type definitions. We then manually annotated the evidence types of 30 evidence items. We tested an RDF/OWL representation of answers to a small number of simple questions about each of these 30 evidence items and showed that automatic inference can determine the detailed evidence types based on this small number of simpler questions. These results show proof-of-concept for a decision support infrastructure that frees the evidence evaluator from mastering relatively complex written evidence type definitions.

The neuronal structure of paramamillary nuclei in Bison bonasus: Nissl and Golgi pictures.

PubMed

Robak, A; Szteyn, S; Równiak, M

1998-01-01

The studies were carried out on the hypothalamus of bison bonasus aged 2 and 3 months. Sections were made by means of Bagiński's technique and Nissl and Klüver-Barrera methods. Four types of neurons were distinguished in the paramamillary nuclei: nucleus supramamillaris (Sm) and nucleus tuberomammillaris pars posterior (Tmp). Type I, small and medium-size, triangular or fusiform cells, which have 2-3 slender, poorly ramified dendrites; typical leptodendritic neurons. Type II, medium size neurons with quadrangular or spindle-shaped perikaryons. Most of them have 3-4 thick dendritic trunks with ramifying relatively long dendrites. These cells show stalked-appearance and possess different appendages sparsely distributed. Type III is similar to type II, but is made of medium-size to large multipolar cells having quadrangular, triangular or fusiform perikaryons and relatively short dendrites. Type IV, small and medium-size, globular cells with 2 or 3 dendritic trunks, which dichotomously subdivide into quaternary dendrites. In all types of neurons, axons emerge from the perikaryon or initial portion of a dendritic trunk. Type I was found in both studied nuclei. Types II and III constitute mainly the nucleus tuberomamillaris pars posterior. Type IV preponderate in the nucleus supramamillaris. The characteristic feature of Tmp cells, in Nissl picture was irregular contour of their somas and clumps of rough Nisls granules, which appear to lie outside the perikaryons. In Sm there were also lightly stained small rounded cells having both small amount of the cytoplasm and tigroid matter.

Identification of two small RNAs within the first 1.5-kb of the herpes simplex virus type 1-encoded latency-associated transcript.

PubMed

Peng, Weiping; Vitvitskaia, Olga; Carpenter, Dale; Wechsler, Steven L; Jones, Clinton

2008-01-01

The herpes simplex virus type 1 (HSV-1) latency-associated transcript (LAT) is abundantly expressed in latently infected neurons. In the rabbit or mouse ocular models of infection, expression of the first 1.5 kb of LAT coding sequences is sufficient for and necessary for wild-type levels of spontaneous reactivation from latency. The antiapoptosis functions of LAT, which maps to the same 1.5 kb of LAT, are important for the latency-reactivation cycle because replacement of LAT with other antiapoptosis genes (the baculovirus IAP gene or the bovine herpesvirus type 1 latency-related gene) restores wild-type levels of reactivation to a LAT null mutant. A recent study identified a micro-RNA within LAT that can inhibit apoptosis (Gupta et al, Nature 442: 82-85). In this study, the authors analyzed the first 1.5 kb of LAT for additional small RNAs that may have regulatory functions. Two LAT-specific small RNAs were detected in productively infected human neuroblastoma cells within the first 1.5 kb of LAT, in a region that is important for inhibiting apoptosis. Although these small RNAs possess extensive secondary structure and a stem-loop structure, bands migrating near 23 bases were not detected suggesting these small RNAs are not true micro-RNAs. Both of the small LAT-specific RNAs have the potential to base pair with the ICP4 mRNA. These two small LAT RNAs may play a role in the latency-reactivation cycle by reducing apoptosis and/or by reducing ICP4 RNA expression.

Differences in healthy food supply and stocking practices between small grocery stores, gas-marts, pharmacies and dollar stores.

PubMed

Caspi, Caitlin Eicher; Pelletier, Jennifer E; Harnack, Lisa; Erickson, Darin J; Laska, Melissa N

2016-02-01

Little is known about the practices for stocking and procuring healthy food in non-traditional food retailers (e.g., gas-marts, pharmacies). The present study aimed to: (i) compare availability of healthy food items across small food store types; and (ii) examine owner/manager perceptions and stocking practices for healthy food across store types. Descriptive analyses were conducted among corner/small grocery stores, gas-marts, pharmacies and dollar stores. Data from store inventories were used to examine availability of twelve healthy food types and an overall healthy food supply score. Interviews with managers assessed stocking practices and profitability. Small stores in Minneapolis and St. Paul, MN, USA, not participating in the Special Supplemental Nutrition Program for Women, Infants, and Children. One hundred and nineteen small food retailers and seventy-one store managers. Availability of specific items varied across store type. Only corner/small grocery stores commonly sold fresh vegetables (63% v. 8% of gas-marts, 0% of dollar stores and 23% of pharmacies). More than half of managers stocking produce relied on cash-and-carry practices to stock fresh fruit (53%) and vegetables (55%), instead of direct store delivery. Most healthy foods were perceived by managers to have at least average profitability. Interventions to improve healthy food offerings in small stores should consider the diverse environments, stocking practices and supply mechanisms of small stores, particularly non-traditional food retailers. Improvements may require technical support, customer engagement and innovative distribution practices.

Differences in healthy food supply and stocking practices between small grocery stores, gas-marts, pharmacies and dollar stores

PubMed Central

Caspi, Caitlin Eicher; Pelletier, Jennifer E.; Harnack, Lisa; Erikson, Darin J.; Laska, Melissa N.

2015-01-01

Objective Little is known about the practices for stocking and procuring healthy food in non-traditional food retailers (e.g., gas-marts, pharmacies). This study aimed to: (i) compare availability of healthy food items across small food store types, and (ii) examine owner/manager perceptions and stocking practices for healthy food across store types. Design Descriptive analyses were conducted among corner/small grocery stores, gas-marts, pharmacies, and dollar stores. Data from store inventories were used to examine availability of 12 healthy food types and an overall healthy food supply score. Interviews with managers assessed stocking practices and profitability. Setting Small stores in Minneapolis and St. Paul, MN not participating in the Special Supplemental Nutrition Program for Women, Infants, and Children (WIC) Subjects 119 small food retailers and 71 store managers Results Availability of specific items varied across store types. Only corner/small grocery stores commonly sold fresh vegetables (63%, versus 8% of food-gas marts, 0% dollar stores, and 23% pharmacies). More than half of managers stocking produce relied on cash and carry practices to stock fresh fruit (53%) and vegetables (55%), instead of direct store delivery. Most healthy foods were perceived by managers to have at least average profitability. Conclusions Interventions to improve healthy food offerings in small stores should consider the diverse environments, stocking practices and supply mechanisms of small stores, particularly non-traditional food retailers. Improvements may require technical support, customer engagement, and innovative distribution practices. PMID:26411535

CECE: A Deep Throttling Demonstrator Cryogenic Engine for NASA's Lunar Lander

NASA Technical Reports Server (NTRS)

Giuliano, Victor J.; Leonard, Timothy G.; Adamski, Walter M.; Kim, Tony S.

2007-01-01

As one of the first technology development programs awarded under NASA's Vision for Space Exploration, the Pratt & Whitney Rocketdyne (PWR) Deep Throttling, Common Extensible Cryogenic Engine (CECE) program was selected by NASA in November 2004 to begin technology development and demonstration toward a deep throttling, cryogenic Lunar Lander engine for use across multiple human and robotic lunar exploration mission segments with extensibility to Mars. The CECE program leverages the maturity and previous investment of a flight-proven hydrogen/oxygen expander cycle engine, the RL10, to develop and demonstrate an unprecedented combination of reliability, safety, durability, throttlability, and restart capabilities in a high-energy, cryogenic engine. NASA Marshall Space Flight Center and NASA Glenn Research Center personnel were integral design and analysis team members throughout the requirements assessment, propellant studies and the deep throttling demonstrator elements of the program. The testbed selected for the initial deep throttling demonstration phase of this program was a minimally modified RL10 engine, allowing for maximum current production engine commonality and extensibility with minimum program cost. In just nine months from technical program start, CECE Demonstrator No. 1 engine testing in April/May 2006 at PWR's E06 test stand successfully demonstrated in excess of 10:1 throttling of the hydrogen/oxygen expander cycle engine. This test provided an early demonstration of a viable, enabling cryogenic propulsion concept with invaluable system-level technology data acquisition toward design and development risk mitigation for both the subsequent CECE Demonstrator No. 2 program and to the future Lunar Lander Design, Development, Test and Evaluation effort.

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

EMERALD REV. 1

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

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

1974-03-01

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

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

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

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

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

NASA Astrophysics Data System (ADS)

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

2012-07-01

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

75 FR 78953 - Department of the Treasury Acquisition Regulation

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-17

...) A small business, women-owned small business, small disadvantaged business, small business owned and... optimal contract types, including considerations of performance based approaches, small business..., including any business or personal information provided, such as names, addresses, e-mail addresses, or...

Kerosene-base fuels in small gasoline engines. Final report

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

Failla, C.C.; Pouring, A.A.

1991-01-01

This document presents the results of an engineering study to demonstrate the technology for converting small gasoline spark-ignited engines, to burn kerosene type fuels to power small generators (0.5 to 3.0 kw). Commercially available (plus those in the developmental stage), reciprocating, two-stroke, four stroke and rotary engines were evaluated for their conversion potential. Unique combustion systems were identified and trade-off studies conducted on engine type, combustion systems, and modification required to burn kerosene type fuels, with special emphasis given to minimizing life cycle cost. Recommendations for the most feasible system are given.

Accuracy of cytology in sub typing non small cell lung carcinomas.

PubMed

Patel, Trupti S; Shah, Majal G; Gandhi, Jahnavi S; Patel, Pratik

2017-07-01

Sub typing of non small cell lung carcinoma (NSCLC) has an important task in the era of molecular and targeted therapies. Differentiating between squamous cell carcinoma (SQCC) and adenocarcinoma (ADC) is challenging when limited material is available in lung carcinoma. We investigated the accuracy and feasibility of sub typing NSCLCs in cytology and small biopsy material. Concurrent cytology and biopsy material obtained in a single CT- guided procedure in lung carcinoma over a year period retrospectively. Both materials were individually sub typed and analyzed. Immunohistochemistry (IHC) was performed. Accuracy was determined by comparing the results with IHC. Total 107 of 126 cases of NSCLCs were included for analysis, where both cytology and biopsy material were adequate for interpretation. FNAC allowed tumor typing in 83 (77.6%) cases; 36 (33.6%) were ADC, 47 (43.9%) cases were SQCC and 24 (22.4%) cases diagnosed as Non-small cell carcinoma not otherwise specified (NSCLC-NOS). In biopsy, 86 cases (80.4%) were typed, among which 34 (31.8%) were ADC, 52 (48.6%) were SQCC and 21 (19.6%) were of NSCLC-NOS type. The result of Chi-square index was significant. With the aid of IHC, NSCLC-NOS reduced from 14 (13%) cases to 2 (1.9%) cases. Cytology and small biopsy specimens achieved comparable specificity and accuracy in sub-typing NSCLC and optimal results were obtain when findings from both modalities combine. The advantage of paired specimens is to maximize overall diagnostic yield and the remaining material will be available for ancillary technique like IHC or for molecular testing. Diagn. Cytopathol. 2017;45:598-603. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

Spatial Distribution of Small Water Body Types across Indiana Ecoregions

EPA Science Inventory

Due to their large numbers and biogeochemical activity, small water bodies (SWB), such as ponds and wetlands, can have substantial cumulative effects on hydrologic, biogeochemical, and biological processes; yet the spatial distributions of various SWB types are often unknown. Usi...

Characteristic pulse trains of preliminary breakdown in four isolated small thunderstorms

NASA Astrophysics Data System (ADS)

Ma, Dong

2017-03-01

Using a low-frequency six-station local network, preliminary breakdown (PB) pulses not followed or followed by negative return stroke (RS), which are defined as PB-type and PB cloud-to-ground (PBCG)-type flashes, are analyzed based on four isolated small thunderstorms for the first time. On the basis of 22 PB-type flashes out of totally 2155 flashes, it indicates that the number of PB-type flashes is very small. At the early stage, PB-type flashes are observed in all four thunderstorms. At the active stage, PB-type flashes still can occur; meanwhile, there are few or no negative cloud-to-ground (CG) flashes. However, at the final stage no PB-type flashes occur. At the stage of distinct cell merging or splitting, PB-type flashes are also observed. Based on the 123 PBCG-type flashes, we discuss the percentage of PBCG-type flashes and also analyze the relationship between the electric field (E-field) amplitude of the largest pulse in the PB pulse train normalized to 100 km (PBA), the E-field amplitude of the first return stroke normalized to 100 km (RSA), the time interval between PBA and RSA (PB-RS interval), and the ratio between PBA and RSA (PB-RS ratio). We find that the percentage of PBCG-type flashes is not always dependent on PBA or PB-RS ratio; the type of thunderstorms may also have an impact on this percentage. None of the PB-RS intervals is less than 20 ms; we speculate that such long PB-RS interval is the feature of isolated small thunderstorms, but more observations are needed to further investigate this question.

Characterization of the interactions of type XII collagen with two small proteoglycans from fetal bovine tendon, decorin and fibromodulin.

PubMed

Font, B; Eichenberger, D; Rosenberg, L M; van der Rest, M

1996-11-01

In addition to the major collagens, such as type I or type II, connective tissues contain a number of less abundant collagens and proteoglycans, whose association contributes to the different properties of the tissues. Type XII and type XIV collagens have been described in soft connective tissues, and type XIV collagen has been shown to interact specifically with decorin through its glycosaminoglycan chain (Font et al., J. Biol. Chem. 268, 25015-25018, 1993). Interactions between these collagens and the small proteoglycans have been characterized further by studying the binding of type XII collagen to decorin by solid phase assays. Our results show a saturable binding of the proteoglycan through its glycosaminoglycan chain to type XII collagen, which does not seem to involve the large non-collagenous NC3 domain of the molecule. This interaction is strongly inhibited by heparin. Furthermore, we report that another small proteoglycan, fibromodulin, isolated from tendon under non-denaturing conditions, is able to bind to type XII collagen. This interaction has been characterized and, unlike that observed with decorin, type XII collagen-fibromodulin interaction seems to take place with the core protein of the proteoglycan. In addition, we report that type XII-type I collagen interactions are not necessarily mediated by decorin as previously suggested.

Classification of small molecule protein kinase inhibitors based upon the structures of their drug-enzyme complexes.

PubMed

Roskoski, Robert

2016-01-01

Because dysregulation and mutations of protein kinases play causal roles in human disease, this family of enzymes has become one of the most important drug targets over the past two decades. The X-ray crystal structures of 21 of the 27 FDA-approved small molecule inhibitors bound to their target protein kinases are depicted in this paper. The structure of the enzyme-bound antagonist complex is used in the classification of these inhibitors. Type I inhibitors bind to the active protein kinase conformation (DFG-Asp in, αC-helix in). Type I½ inhibitors bind to a DFG-Asp in inactive conformation while Type II inhibitors bind to a DFG-Asp out inactive conformation. Type I, I½, and type II inhibitors occupy part of the adenine binding pocket and form hydrogen bonds with the hinge region connecting the small and large lobes of the enzyme. Type III inhibitors bind next to the ATP-binding pocket and type IV inhibitors do not bind to the ATP or peptide substrate binding sites. Type III and IV inhibitors are allosteric in nature. Type V inhibitors bind to two different regions of the protein kinase domain and are therefore bivalent inhibitors. The type I-V inhibitors are reversible. In contrast, type VI inhibitors bind covalently to their target enzyme. Type I, I½, and II inhibitors are divided into A and B subtypes. The type A inhibitors bind in the front cleft, the back cleft, and near the gatekeeper residue, all of which occur within the region separating the small and large lobes of the protein kinase. The type B inhibitors bind in the front cleft and gate area but do not extend into the back cleft. An analysis of the limited available data indicates that type A inhibitors have a long residence time (minutes to hours) while the type B inhibitors have a short residence time (seconds to minutes). The catalytic spine includes residues from the small and large lobes and interacts with the adenine ring of ATP. Nearly all of the approved protein kinase inhibitors occupy the adenine-binding pocket; thus it is not surprising that these inhibitors interact with nearby catalytic spine (CS) residues. Moreover, a significant number of approved drugs also interact with regulatory spine (RS) residues. Copyright © 2015 Elsevier Ltd. All rights reserved.

Human tRNA-derived small RNAs in the global regulation of RNA silencing

PubMed Central

Haussecker, Dirk; Huang, Yong; Lau, Ashley; Parameswaran, Poornima; Fire, Andrew Z.; Kay, Mark A.

2010-01-01

Competition between mammalian RNAi-related gene silencing pathways is well documented. It is therefore important to identify all classes of small RNAs to determine their relationship with RNAi and how they affect each other functionally. Here, we identify two types of 5′-phosphate, 3′-hydroxylated human tRNA-derived small RNAs (tsRNAs). tsRNAs differ from microRNAs in being essentially restricted to the cytoplasm and in associating with Argonaute proteins, but not MOV10. The first type belongs to a previously predicted Dicer-dependent class of small RNAs that we find can modestly down-regulate target genes in trans. The 5′ end of type II tsRNA was generated by RNaseZ cleavage downstream from a tRNA gene, while the 3′ end resulted from transcription termination by RNA polymerase III. Consistent with their preferential association with the nonslicing Argonautes 3 and 4, canonical gene silencing activity was not observed for type II tsRNAs. The addition, however, of an oligonucleotide that was sense to the reporter gene, but antisense to an overexpressed version of the type II tsRNA, triggered robust, >80% gene silencing. This correlated with the redirection of the thus reconstituted fully duplexed double-stranded RNA into Argonaute 2, whereas Argonautes 3 and 4 were skewed toward less structured small RNAs, particularly single-strand RNAs. We observed that the modulation of tsRNA levels had minor effects on the abundance of microRNAs, but more pronounced changes in the silencing activities of both microRNAs and siRNAs. These findings support that tsRNAs are involved in the global control of small RNA silencing through differential Argonaute association, suggesting that small RNA-mediated gene regulation may be even more finely regulated than previously realized. PMID:20181738

Approach to numerical safety guidelines based on a core melt criterion. [PWR; BWR

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

Azarm, M.A.; Hall, R.E.

1982-01-01

A plausible approach is proposed for translating a single level criterion to a set of numerical guidelines. The criterion for core melt probability is used to set numerical guidelines for various core melt sequences, systems and component unavailabilities. These guidelines can be used as a means for making decisions regarding the necessity for replacing a component or improving part of a safety system. This approach is applied to estimate a set of numerical guidelines for various sequences of core melts that are analyzed in Reactor Safety Study for the Peach Bottom Nuclear Power Plant.

Spent fuel data base: commercial light water reactors. [PWR; BWR

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

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Material control and accountancy at EDF PWR plants; GCN: Gestion du Combustible Nucleaire

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

de Cormis, F.

1991-01-01

The paper describes the comprehensive system which is developed and implemented at Electricite de France to provide a single reliable nuclear material control and accounting system for all nuclear plants. This software aims at several objectives among which are: the control and the accountancy of nuclear material at the plant, the optimization of the consistency of data by minimizing the possibility of transcription errors, the fulfillment of the statutory requirements by automatic transfer of reports to national and international safeguards authorities, the servicing of other EDF users of nuclear material data for technical or commercial purposes.

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

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times.

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

DTIC Science & Technology

2009-12-01

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

Pinhole-type two-dimensional ultra-small-angle X-ray scattering on the micrometer scale

PubMed Central

Kishimoto, Hiroyuki; Shinohara, Yuya; Suzuki, Yoshio; Takeuchi, Akihisa; Yagi, Naoto; Amemiya, Yoshiyuki

2014-01-01

A pinhole-type two-dimensional ultra-small-angle X-ray scattering set-up at a so-called medium-length beamline at SPring-8 is reported. A long sample-to-detector distance, 160.5 m, can be used at this beamline and a small-angle resolution of 0.25 µm−1 was thereby achieved at an X-ray energy of 8 keV. PMID:24365910

From Paper to Production to Test: An Update on NASA's J-2X Engine for Exploration

NASA Technical Reports Server (NTRS)

Kynard, Michael

2011-01-01

The NASA/industry team responsible for developing the J-2X upper stage engine for the Space Launch System (SLS) Program has made significant progress toward moving beyond the design phase and into production, assembly, and test of development hardware. The J-2X engine exemplifies the SLS Program goal of using proven technology and experience from more than 50 years of United States spaceflight experience combined with modern manufacturing processes and approaches. It will power the second stage of the fully evolved SLS Program launch vehicle that will enable a return to human exploration of space beyond low earth orbit. Pratt & Whitney Rocketdyne (PWR) is under contract to develop and produce the engine, leveraging its flight-proven LH2/LOX, gas generator cycle J-2 and RS-68 engine capabilities, recent experience with the X-33 aerospike XRS-2200 engine, and development knowledge of the J-2S tap-off cycle engine. The J- 2X employs a gas generator operating cycle designed to produce 294,000 pounds of vacuum thrust in primary operating mode with its full nozzle extension. With a truncated nozzle extension suitable to support engine clustering on the stage, the nominal vacuum thrust level in primary mode is 285,000 pounds. It also has a secondary mode, during which it operates at 80 percent thrust by altering its mixture ratio. The J-2X development philosophy is based on proven hardware, an aggressive development schedule, and early risk reduction. NASA Marshall Space Flight Center (MSFC) and PWR began development of the J-2X in June 2006. The government/industry team of more than 600 people within NASA and PWR successfully completed the Critical Design Review (CDR) in November 2008, following extensive risk mitigation testing. Assembly of the first development engine was completed in May 2011 and the first engine test was conducted at the NASA Stennis Space Center (SSC), test stand A2, on 14 July 2011. Testing of the first development engine will continue through the autumn of 2011, be paused for test stand modifications to the passive diffuser, and then restart in the spring of 2012. This testing will be followed by specialized powerpack testing intended to examine the design and operating margins of the engine turbomachinery. The development plan beyond this point leads through more system-level, engine testing of several samples, analytical model validation activities, functional and performance verification, and then ultimate certification to support human spaceflight. This paper will discuss the J-2X development background, provide top-level information on design and development planning, and will explore some of the development challenges and mitigation activities pursued to date.

48 CFR 19.705-4 - Reviewing the subcontracting plan.

Code of Federal Regulations, 2011 CFR

2011-10-01

... disadvantaged business, and women-owned small business subcontractors to the maximum practicable extent. The..., service-disabled veteran-owned small business, HUBZone small business, or women-owned small business... women-owned small business concerns. If information is not available on a specific type of product or...

48 CFR 19.705-4 - Reviewing the subcontracting plan.

Code of Federal Regulations, 2010 CFR

2010-10-01

... disadvantaged business, and women-owned small business subcontractors to the maximum practicable extent. The..., service-disabled veteran-owned small business, HUBZone small business, or women-owned small business... women-owned small business concerns. If information is not available on a specific type of product or...

In Situ Oxidation Synthesis of p-Type Composite with Narrow-Bandgap Small Organic Molecule Coating on Single-Walled Carbon Nanotube: Flexible Film and Thermoelectric Performance.

PubMed

Gao, Caiyan; Chen, Guangming

2018-03-01

Although composites of organic polymers or n-type small molecule/carbon nanotube (CNT) have achieved significant advances in thermoelectric (TE) applications, p-type TE composites of small organic molecules as thick surface coating layers on the surfaces of inorganic nanoparticles still remain a great challenge. Taking advantage of in situ oxidation reaction of thieno[3,4-b]pyrazine (TP) into TP di-N-oxide (TPNO) on single-walled CNT (SWCNT) surface, a novel synthesis strategy is proposed to achieve flexible films of TE composites with narrow-bandgap (1.19 eV) small molecule coating on SWCNT surface. The TE performance can be effectively enhanced and conveniently tuned by poly(sodium-p-styrenesulfonate) content, TPNO/SWCNT mass ratio, and posttreatment by various polar solvents. The maximum of the composite power factor at room temperature is 29.4 ± 1.0 µW m -1 K -2 . The work presents a way to achieve flexible films of p-type small organic molecule/inorganic composites with clear surface coating morphology for TE application. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Small mammal habitat associations in poletimber and sawtimber stands of four forest cover types

Treesearch

Richard M. DeGraaf; Dana P. Snyder; Barbara J. Hill

1991-01-01

Small mammal distribution was examined in poletimber and sawtimber stands of four forest cover types in northern New England: northern hardwoods, red maple, balsam fir, and red spruce-balsam fir. During 1980 and 1981, eight stands on the White Mountain National Forest, NH, were sampled with four trap types (three sizes of snap traps and one pit-fall) for 16 000 trap-...

13 CFR 107.1100 - Types of Leverage and application procedures.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Types of Leverage and application procedures. 107.1100 Section 107.1100 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION SMALL BUSINESS INVESTMENT COMPANIES SBA Financial Assistance for Licensees (Leverage) General Information About...

Gamma ray-induced small plaque mutants of western equine encephalitis virus

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

Simizu, B.; Yamazaki, S.; Suzuki, K.

1973-12-01

Small plaque mutants of Western equine encephalitis virus were obtained from the surviving fractions of wild-type virus which was irradiated with gamma rays. The frequency with which small plaque mutants appeared in the surviving fraction increased with the radiation dose. These mutants were not more resistant to radiation than wild-type virus. The growth rate of a mutant, S127, was lower than that of wild-type. Clonally purified mutant virions presented two peaks in a velocity sedimentation profile; peak 1 corresponded to the peak of wild type and peak 2 moved faster than peak 1. Virions of both peaks were infectious andmore » consistently formed small plaques in chicken embryo cells. Virions reisolated from either peak and grown in chicken embryo cells also revealed two peaks in sedimentation analysis. In the electron microscope examination peak 2 proved to consist of giant form particles, each of which contained more than one nucleoid surrounded with a common envelope. Despite this remarkable morphological difference, densities of the wild-type and S127 mutant virions were similar in cesium chloride gradients. The RNAs and proteins of mutant virions could not be distinguished from those of wild types on the basis of size or change. (auth)« less

Longitudinal Monitoring of Successive Commercial Layer Flocks for Salmonella enterica Serovar Enteritidis.

PubMed

Denagamage, Thomas N; Patterson, Paul; Wallner-Pendleton, Eva; Trampel, Darrell; Shariat, Nikki; Dudley, Edward G; Jayarao, Bhushan M; Kariyawasam, Subhashinie

2016-11-01

The Pennsylvania Egg Quality Assurance Program (EQAP) provided the framework for Salmonella Enteritidis (SE) control programs, including the Food and Drug Administration (FDA) mandated Final Egg Rule, for commercial layer facilities throughout the United States. Although flocks with ≥3000 birds must comply with the FDA Final Egg Rule, smaller flocks are exempted from the rule. As a result, eggs produced by small layer flocks may pose a greater public health risk than those from larger flocks. It is also unknown if the EQAPs developed with large flocks in mind are suitable for small- and medium-sized flocks. Therefore, a study was performed to evaluate the effectiveness of best management practices included in EQAPs in reducing SE contamination of small- and medium-sized flocks by longitudinal monitoring of their environment and eggs. A total of 59 medium-sized (3000 to 50,000 birds) and small-sized (<3000 birds) flocks from two major layer production states of the United States were enrolled and monitored for SE by culturing different types of environmental samples and shell eggs for two consecutive flock cycles. Isolated SE was characterized by phage typing, pulsed-field gel electrophoresis (PFGE), and clustered regularly interspaced short palindromic repeats-multi-virulence-locus sequence typing (CRISPR-MVLST). Fifty-four Salmonella isolates belonging to 17 serovars, 22 of which were SE, were isolated from multiple sample types. Typing revealed that SE isolates belonged to three phage types (PTs), three PFGE fingerprint patterns, and three CRISPR-MVLST SE Sequence Types (ESTs). The PT8 and JEGX01.0004 PFGE pattern, the most predominant SE types associated with foodborne illness in the United States, were represented by a majority (91%) of SE. Of the three ESTs observed, 85% SE were typed as EST4. The proportion of SE-positive hen house environment during flock cycle 2 was significantly less than the flock cycle 1, demonstrating that current EQAP practices were effective in reducing SE contamination of medium and small layer flocks.

What type of rural? Assessing the variations in life expectancy at birth at small area-level for a small population province using classes of locally defined settlement types.

PubMed

Terashima, Mikiko; Read Guernsey, Judith; Andreou, Pantelis

2014-02-13

Although efforts have been made to articulate rural-urban health inequalities in recent years, results have been inconsistent due to different geographical scales used in these studies. Small-area level investigations of health inequalities will likely show more detailed pictures of health inequalities among diverse rural communities, but they are difficult to conduct, particularly in a small population region. The objectives of this study were: 1) to compare life expectancy at birth for females and males across small-areas classified by locally defined settlement types for a small province in Canada; 2) to assess whether any of the settlement types explains variations in life expectancy over and above the extent of socioeconomic disadvantage and social isolation; and 3) to examine variations in life expectancies within a (larger) area unit used as the basis of health inequality investigations in previous studies. Seven settlement types were determined for the 'community' units based on population per-kilometre-road density and settlement forms. Mean life expectancies at birth for both genders were compared by settlement type, both for the entire province and within the Halifax Regional Municipality--the province's only census designated metropolitan area, but also contains rural settlements. Linear regression analyses were conducted to assess the statistical associations between life expectancy and the settlement types, adjusting for indicators of community-level deprivation. While types of communities considered as 'rural' generally had lower life expectancy for both genders, the effects of living in any settlement type were attenuated once adjusted for socioeconomic deprivation and social isolation. An exception was the village and settlement cluster type, which had additionally negative effects on health for females. There were some variations observed within the Halifax Regional Municipality, suggesting the importance of further investigating a variety of health and disease outcomes at smaller area-levels than those employed in previous studies. This paper highlighted the importance of further articulating the differences in the characteristics of rural at finer area-levels and the differential influence they may have on health. Further efforts are desirable to overcome various data challenges in order to extend the investigation of health inequalities to hard-to-study provinces.

[What is the prognostic significance of histomorphology in small cell lung carcinoma?].

PubMed

Facilone, F; Cimmino, A; Assennato, G; Sardelli, P; Colucci, G A; Resta, L

1993-01-01

What is the prognostic significant of the histomorphology in the small cell carcinomas of the lung? After the WHO classification of the lung cancer (1981), several studies criticized the subdivision of the small cell carcinoma in three sub-types (oat-cell, intermediate cell and combined types). The role of histology in the prognostic predition has been devaluated. In order to verify the prognostic value of the morphology of the small cell types of lung cancer, we performed a multivariate analysis in 62 patients. The survival rate was analytically compared with the following parameters: nuclear maximum diameter, nuclear form, nuclear chromatism, chromatine distribution, presence of nucleolus, evidence of cytoplasm. The results showed that none of these parameters are able to express a prognostic value. According to the recent studies, we think that the small cell carcinoma of the lung is a neoplasia with a multiform histologic pattern. Differences observed in clinical management are not correlate with the morphology, but with other biological parameters still unknown.

Evaluation of Visual Alerts in the Maritime Domain. Study 2. Program Modifications

DTIC Science & Technology

2009-02-01

feedback that they were wrong, and without consulting the Status screen again enter the alternate answer (“ qwe ”). That is, the need to consult the...Neutral Large Slow No QWE Hostile Small Fast Yes ASD DRDC Atlantic CR 2008-268 5 Table 3. First proposed target types and...classification scheme. Target Size Speed Weapons Flag Response Type Neutral Large Slow No Other QWE Hostile Small Fast Yes Other ASD Friendly Large/Small Slow

Wilderness ecology: the upland plant communities, woody browse production, and small mammals of two adjacent 33-year-old wildfire areas of northeastern Minnesota.

Treesearch

Lewis F. Ohmann; Charles T. Cushwa; Roger E. Lake; James R. Beer; Robert B. Brander

1973-01-01

In three parts, classifies the upland vegetation into four community types; describes the measurements of browse and gives yields by different species; and describes the relation of small mammal populations to vegetative community types.

Fundamental molecular design for precise control of thermoresponsiveness of organic polymers by using ternary systems.

PubMed

Amemori, Shogo; Kokado, Kenta; Sada, Kazuki

2012-05-23

The de novo design of thermosensitive polymers in solution has been achieved by using the addition of small organic molecules (or "effectors"). Hydrogen bonding as an attractive polymer-polymer or polymer-effector interaction substantially dominates the responsivity, causing facile switching between LCST-type and UCST-type phase transitions, control of the transition temperature, and further coincidence of the two transitions. Small molecules having a high affinity for the polymer induce UCST-type phase behavior, whereas those having a low affinity for the polymer showed LCST-type phase behavior.

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

Gerard, R.; Malekian, C.; Meessen, O.

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports inmore » the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.« less

Isoniazid interaction with phosphatidylcholine-based membranes

NASA Astrophysics Data System (ADS)

Marques, Amanda Vicente; Marengo Trindade, Paulo; Marques, Sheylla; Brum, Tainá; Harte, Etienne; Rodrigues, Marieli Oliveira; D'Oca, Marcelo Gonçalves Montes; da Silva, Pedro Almeida; Pohlmann, Adriana R.; Alves, Isabel Dantas; de Lima, Vânia Rodrigues

2013-11-01

Interaction between the anti-tuberculosis drug isoniazid (INH) and phosphatidylcholine membranes was investigated in terms of: (i) drug affinity to a lipid bilayer and (ii) drug-induced changes in the dynamic properties of liposomes, such as membrane hydration state, polar head and non-polar acyl chain order and lipid phase transition behavior. These parameters were studied by plasmon waveguide resonance spectroscopy (PWR), UV-visible, horizontal attenuated total reflectance-Fourier transform infrared (HATR-FTIR), nuclear magnetic resonance (NMR) and differential scanning calorimetry (DSC) techniques. PWR measurements showed an INH membrane dissociation constant value of 0.031 μM to phosphatidylcholine bilayers. INH induced higher membrane perturbation in the plane which is perpendicular to the membrane plane. The INH saturation concentration in phosphatidylcholine liposomes was 170 μM. At this concentration, HATR-FTIR and NMR findings showed that INH may interact with the lipid polar head, increasing the number of hydrogen bonds in the phosphate region and enhancing the choline motional freedom. DSC measurements showed that, at 115 μM, INH was responsible for a decrease in lipid phase transition temperature of approximately 2 °C and had no influence in the lipid enthalpy variation (ΔH). However, at 170 μM, INH induced the reduction of the ΔH by approximately 52%, suggesting that the drug may increase the distance among lipid molecules and enhance the freedom of the lipid acyl chains methylene groups. This paper provides information on the effects of INH on membrane dynamics which is important to understand liposome targeting of the drug and for the development of anti-TB pharmacologic systems that not only are less susceptible to resistance but also have low toxicity.

Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

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

Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.

2012-07-01

The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now.more » The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)« less

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

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

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

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

2015-09-01

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

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

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

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

1989-03-01

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

J-2X Fuel Turbopump Point of Departure: The Performance of the J-2s Fuel Turbopump Inducer

NASA Technical Reports Server (NTRS)

Sargent, S. R.; Becht, D. G.; Mulder, A. D.

2008-01-01

To aid the J-2X program design effort with detailed performance and environment information, the J-2S fuel turbopump (FTP) inducer has undergone a thorough test series in both water and hydrogen. The test series utilizes both inducer only and a complete pump configuration to assess the inducer interaction to the overall turbopump system. The test goals include verification of suction performance against heritage J-2S data, head production, effects of thermodynamic suppression head (TSH), and evaluation of the inducer dynamic pressure caused by cavitation instabilities. Test facilities at both Pratt & Whitney Rocketdyne (PWR) and NASA s Stennis Space Center (SSC) are employed for the testing. The inducer only water test effort conducted at PWR established performance curves for suction performance, head production, and efficiency over a wide operating range. Because the heritage J-2S suction performance data set is in hydrogen, it is desired to obtain current suction performance data in hydrogen as well, thus avoiding the reliance on a theoretical TSH correlation for direct comparison. This data then provides an empirically based TSH correlation allowing for the comparison of water test suction data to system suction requirements. The FTP testing performed at SSC provides these suction performance relationships as well as inlet duct dynamic pressures during liquid hydrogen operation. The test effort successfully confirms the heritage J-2S suction performance and establishes the amount of TSH between water and hydrogen operation at the design flow coefficient. Correlating data is also obtained for cavitating instability frequency content, illustrating the validity of using the wide flow range water test data to predict hydrogen performance.

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

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

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

2011-01-01

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

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

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.

A Study of Cavitation Erosion

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

Hiromu Isaka; Masatsugu Tsutsumi; Tadashi Shiraishi

2002-07-01

The authors performed experimental study for the purpose of the following two items from a viewpoint of cavitation erosion of a cylindrical orifice in view of a problem at the letdown orifice in PWR (Pressurized Water Reactor). 1. To get the critical cavitation parameter of the cylindrical orifice to establish the design criteria for prevention of cavitation erosion, and 2. to ascertain the erosion rate in such an eventuality that the cavitation erosion occurs with the orifice made of stainless steel with precipitation hardening (17-4-Cu hardening type stainless steel), so that we confirm the appropriateness of the design criteria. Regardingmore » the 1. item, we carried out the cavitation tests to get the critical cavitation parameters inside and downstream of the orifice. The test results showed that the cavitation parameter at inception is independent of the length or the diameter of the orifice. Moreover, the design criteria of cavitation erosion of cylindrical orifices have been established. Regarding the 2. item, we tested the erosion rate under high-pressure conditions. The cavitation erosion actually occurred in the cylindrical orifice at the tests that was strongly resemble to the erosion occurred at the plant. It will be seldom to reproduce resemble cavitation erosion in a cylindrical orifice with the hard material used at plants. We could establish the criteria for preventing the cavitation erosion from the test results. (authors)« less

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

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

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

A comparison study of different RF shields for an 8-element transceive small animal array at 9.4T.

PubMed

Jin, Jin; Li, Yu; Liu, Feng; Weber, Ewald; Crozier, Stuart

2011-01-01

In this study, three types of radio-frequency shields are studied and compared in the context of ultra-high field small-animal magnetic resonance imaging. It has been demonstrated that the coil penetration depth and mutual coupling between the coils depend heavily on the type of shield employed. The results were used to guide the design of a 9.4T 8-element transceive small animal array, which provides high overall coil penetration.

13 CFR 107.825 - Purchasing securities from an underwriter or other third party.

Code of Federal Regulations, 2010 CFR

2010-01-01

... ADMINISTRATION SMALL BUSINESS INVESTMENT COMPANIES Financing of Small Businesses by Licensees Structuring Licensee's Financing of Eligible Small Businesses: Types of Financing § 107.825 Purchasing securities from...

Student Perceptions of Small-Group Learning

ERIC Educational Resources Information Center

Florez, Ida Rose; McCaslin, Mary

2008-01-01

Background/Context: Elementary school teachers regularly arrange students in small groups for learning activities. A rich literature discusses various types of small-group learning formats and how those formats affect achievement. Few studies, however, have examined students' perceptions of small-group learning experiences. Our work extends the…

Effects of icotinib on advanced non-small cell lung cancer with different EGFR phenotypes.

PubMed

Pan, Huiyun; Liu, Rong; Li, Shengjie; Fang, Hui; Wang, Ziwei; Huang, Sheng; Zhou, Jianying

2014-09-01

Icotinib is the first oral epidermal growth factor receptor (EGFR) tyrosine kinase receptor inhibitor, which has been proven to exert significant inhibitory effects on non-small cell lung cancer in vitro. Clinical evidence has showed that the efficacy of Icotinib on retreating advanced non-small cell lung cancer is comparable to Gefitinib. However, different phenotypes of EGFR can affect the therapeutic outcomes of EGFR tyrosine kinase receptor inhibitor. Therefore, our study focused on efficacy and safety of Icotinib in patients with advanced non-small cell lung cancer of different EGPR phenotypes. Clinical data of patients with advanced non-small cell lung cancer who received Icotinib treatment from August, 2011 to May, 2013 were retrospectively analyzed. Kaplan-Meier analysis was used for survival analysis and comparison. 18 wild-type EGFR and 51 mutant type were found in a total of 69 patients. Objective response rate of patients with mutant type EGFR was 54.9 % and disease control rate was 86.3 %. Objective response rate of wild-type patients was 11.1 % (P = 0.0013 vs mutant type), disease control rate was 50.0 % (P = 0.0017). Median progression-free survival (PFS) of mutant type and wild-type patients were 9.7 and 2.6 months, respectively (P < 0.001). Median PFS of exon 19 mutated mutant patients was 11.3 months, mean PFS of exon 21 L858R mutated mutant patients was 8.7 months (P = 0.3145). Median overall survival (OS) of EGFR mutated patients had not reached. OS time of 13 wild-type patients was 12.9 months (P < 0.001). The common adverse reactions of Icotinib included rash, diarrhea, itching skin with occurrence rates of 24.6 % (17/69), 13.0 % (9/69), and 11.6 % (8/69), respectively. Most adverse reactions were grade I-II. Icotinib has great efficacy in EGFR mutated patients, making it an optimal regimen to treat EGFR mutated patients. Furthermore, most of adverse reactions associated with Icotinib treatment were tolerable.

[The value of immunohistochemistry using SMARCA4/BRG1 in the diagnosis of small cell ovarian carcinoma hypercalcemic type. A report of two cases].

PubMed

Ruiz-García, Gema; Torroba-Carón, M Amparo; Ferri-Ñíguez, Belén; Lencina-Guardiola, Miriam; García-Molina, Francisco; Martínez-Barba, Enrique

Small cell carcinoma of ovary-hypercalcemic type is an undifferentiated carcinoma. We describe two cases in women aged 32 and 29. Both presented with large masses and complete surgical extirpation was impossible. Histologically, the images were similar, with diffuse cell proliferation, accompanied by the presence of follicle-like spaces. In both cases it was necessary to make a differential diagnosis with entities such as adult or juvenile granulosa cell tumour, small cell carcinoma of pulmonary type, dysgerminoma and even peripheral neuroectodermal tumour. The absence of SMARCA4/BRG1 immunostaining proved very useful in the diagnosis of hypercalcemic small cell ovarian carcinoma. Copyright © 2017 Sociedad Española de Anatomía Patológica. Publicado por Elsevier España, S.L.U. All rights reserved.

A Geology Sampling System for Small Bodies

NASA Technical Reports Server (NTRS)

Hood, A. D.; Naids, A. J.; Graff, T.; Abell, P.

2015-01-01

Human exploration of Small Bodies is being investigated as a precursor to a Mars surface mission. Asteroids, comets, dwarf planets, and the moons of Mars all fall into this Small Bodies category and some are being discussed as potential mission tar-gets. Obtaining geological samples for return to Earth will be a major objective for any mission to a Small Body. Currently the knowledge base for geology sampling in microgravity is in its infancy. Furthermore, humans interacting with non-engineered surfaces in a microgravity environment poses unique challenges. In preparation for such missions, a team at the National Aeronautics and Space Administration (NASA) John-son Space Center (JSC) has been working to gain experience on how to safely obtain numerous sample types in such an environment. This abstract briefly summarizes the type of samples the science community is interested in, discusses an integrated geology sampling solution, and highlights some of the unique challenges associated with this type of exploration.

13 CFR 127.101 - What type of assistance is available under this part?

Code of Federal Regulations, 2010 CFR

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false What type of assistance is available under this part? 127.101 Section 127.101 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION WOMEN-OWNED SMALL BUSINESS FEDERAL CONTRACT ASSISTANCE PROCEDURES General Provisions § 127.101...

Small Learning Communities Sense of Belonging to Reach At-Risk Students of Promise

ERIC Educational Resources Information Center

Hackney, Debbie

2011-01-01

The research design is a quantitative causal comparative method. The Florida Comprehensive Assessment Test (FCAT) which measures student scores included assessments in mathematics and reading. The design study called for an examination of how type of small learning community (SLC) or the type non-SLC high school environment affected student…

Transect versus grid trapping arrangements for sampling small-mammal communities

Treesearch

Dean E. Pearson; Leonard F. Ruggiero

2003-01-01

We compared transect and grid trapping arrangements for assessing small-mammal community composition and relative abundance for 2 years in 2 forest cover types in west-central Montana, USA. Transect arrangements yielded more total captures, more individual captures, and more species than grid arrangements in both cover types in both years. Differences between...

Birth order in small multihospital systems.

PubMed

Luke, R D; Ozcan, Y A; Begun, J W

1990-06-01

The strategic behaviors of small multihospital systems have received little attention in the literature despite the fact that small systems are the predominant scale among multihospital systems. This study examines one important aspect of small-system strategic behaviors: the birth-order or evolutionary patterns of hospital acquisition. The evolutionary patterns of acquisition are compared across three strategic model types studied elsewhere: local market, investment, and historical. Using data obtained from a variety of sources, local market model systems are found, in the sequence of acquisition, to be significantly different from the other two model types in terms of relative distances of acquisitions from the initiating or parent hospital, the sizes of acquisition hospitals, the complexity of those hospitals, and the likelihood that the acquisitions are located in rural areas. Differences between parents and acquisitions are also significant, as hypothesized, for the market model system types, although they are not generally significant for the other two model types. The findings suggest that the market model represents an important strategic form that may have important implications for the restructuring of hospital markets.

Development of a high-sensitivity and portable cell using Helmholtz resonance for noninvasive blood glucose-level measurement based on photoacoustic spectroscopy.

PubMed

Tachibana, K; Okada, K; Kobayashi, R; Ishihara, Y

2016-08-01

We describe the possibility of high-sensitivity noninvasive blood glucose measurement based on photoacoustic spectroscopy (PAS). The demand for noninvasive blood glucose-level measurement has increased due to the explosive increase in diabetic patients. We have developed a noninvasive blood glucose-level measurement based on PAS. The conventional method uses a straight-type resonant cell. However, the cell volume is large, which results in a low detection sensitivity and difficult portability. In this paper, a small-sized Helmholtz-type resonant cell is proposed to improve detection sensitivity and portability by reducing the cell dead volume. First, the acoustic property of the small-sized Helmholtz-type resonant cell was evaluated by performing an experiment using a silicone rubber. As a result, the detection sensitivity of the small-sized Helmholtz-type resonant cell was approximately two times larger than that of the conventional straight-type resonant cell. In addition, the inside volume was approximately 30 times smaller. Second, the detection limits of glucose concentration were estimated by performing an experiment using glucose solutions. The experimental results showed that a glucose concentration of approximately 1% was detected by the small-sized Helmholtz-type resonant cell. Although these results on the sensitivity of blood glucose-level measurement are currently insufficient, they suggest that miniaturization of a resonance cell is effective in the application of noninvasive blood glucose-level measurement.

Towards biodiversity hotspots effective for conserving mammals with small geographic ranges

NASA Astrophysics Data System (ADS)

Carrara, Rodolfo; San Blas, Germán; Agrain, Federico; Roig-Juñent, Sergio

2017-01-01

The main goal of using global biodiversity hotspots for conservation purposes is to protect taxa with small geographic ranges because these are highly vulnerable to extinction. However, the extent to what different hotspots types are effective for meeting this goal remains controversial because hotspots have been previously defined as either the richest or most threatened and richest sites in terms of total, endemic or threatened species. In this regard, the use of species richness to set conservation priorities is widely discussed because strategies focused on this diversity measure tend to miss many of the taxa with small geographic ranges. Here we use data on global terrestrial mammal distributions to show that, hotspots of total species, endemism and threat defined in terms of species richness are effective in including 27%, 29% and 11% respectively, of the taxa with small geographic ranges. Whilst, the same hotspot types defined in terms of a simple diversity index, which is a function of species richness and range-size rarity, include 68%, 44% and 90% respectively, of these taxa. In addition, we demonstrate that index hotspot types are highly efficient because they conserve 79% of mammal species (21% more species than richness hotspot types), with 59% of species shared by three hotspot types (31% more than richness hotspot types). These results suggest that selection of different diversity measures to define hotspots may strongly affect the achievement of conservation goals.

Induction and reversal of myotonic dystrophy type 1 pre-mRNA splicing defects by small molecules.

PubMed

Childs-Disney, Jessica L; Stepniak-Konieczna, Ewa; Tran, Tuan; Yildirim, Ilyas; Park, HaJeung; Chen, Catherine Z; Hoskins, Jason; Southall, Noel; Marugan, Juan J; Patnaik, Samarjit; Zheng, Wei; Austin, Chris P; Schatz, George C; Sobczak, Krzysztof; Thornton, Charles A; Disney, Matthew D

2013-01-01

The ability to control pre-mRNA splicing with small molecules could facilitate the development of therapeutics or cell-based circuits that control gene function. Myotonic dystrophy type 1 is caused by the dysregulation of alternative pre-mRNA splicing due to sequestration of muscleblind-like 1 protein (MBNL1) by expanded, non-coding r(CUG) repeats (r(CUG)(exp)). Here we report two small molecules that induce or ameliorate alternative splicing dysregulation. A thiophene-containing small molecule (1) inhibits the interaction of MBNL1 with its natural pre-mRNA substrates. Compound (2), a substituted naphthyridine, binds r(CUG)(exp) and displaces MBNL1. Structural models show that 1 binds MBNL1 in the Zn-finger domain and that 2 interacts with UU loops in r(CUG)(exp). This study provides a structural framework for small molecules that target MBNL1 by mimicking r(CUG)(exp) and shows that targeting MBNL1 causes dysregulation of alternative splicing, suggesting that MBNL1 is thus not a suitable therapeutic target for the treatment of myotonic dystrophy type 1.

Note: Development of a small maglev-type antirolling system.

PubMed

Park, Cheol Hoon; Park, Hee Chang; Cho, Han Wook; Moon, Seok Jun; Chung, Tae Young

2010-05-01

Various passive and/or active antirolling devices have been used for suppressing the rolling motion of ships in the ocean. In this study, a maglev-type active mass driver (AMD) is developed for controlling the rolling motion of a shiplike structure. No friction is generated during the motion of this maglev-type AMD, as the moving mass is floated by the magnetic levitation force and displaced by the propulsion force generated by the linear motor. For verifying the feasibility of the proposed method, a small AMD having a moving mass of approximately 4.0 kg is constructed and used in a small-scale model of a catamaran. This paper presents the detailed design procedures and obtained experimental results. Our results show that the developed maglev-type AMD has the potential for use in controlling the rolling motion of ships and other oceanographic vessels.

48 CFR 215.203-70 - Requests for proposals-tiered evaluation of offers.

Code of Federal Regulations, 2010 CFR

2010-10-01

... ACQUISITION REGULATIONS SYSTEM, DEPARTMENT OF DEFENSE CONTRACTING METHODS AND CONTRACT TYPES CONTRACTING BY... shall be consistent with FAR part 19. (b) Consideration shall be given to the tiers of small businesses (e.g., 8(a), HUBZone small business, service-disabled veteran-owned small business, small business...

Education and National Scale: The World of Small States.

ERIC Educational Resources Information Center

Brock, Colin

1988-01-01

Analyzes smallness of national scale, types and degrees of isolation, and degrees of dependence as factors influencing the establishment of education in small states. Notes that the context created by these three circumstances makes dependency inevitable. Examines regionalism as a possible solution to the educational problems of small nations. (KO)

Geographical and Geodemographic Connections between Different Types of Small Area as the Origins and Destinations of Migrants to Mid-Wales

ERIC Educational Resources Information Center

Walford, Nigel

2007-01-01

Exchanges of population between supposedly "urban" and "rural" spaces have occurred throughout history as people migrate between areas with relatively, densely and sparsely settled populations. However, comparatively little is known about whether the same small areas persistently contribute to the flow and what types of…

13 CFR 124.702 - What types of assistance are available through the 7(j) program?

Code of Federal Regulations, 2010 CFR

2010-01-01

..., marketing, and operation of small business concerns; and (b) The identification and development of new... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false What types of assistance are available through the 7(j) program? 124.702 Section 124.702 Business Credit and Assistance SMALL BUSINESS...

Posttest TRAC-PD2/MOD1 predictions for FLECHT SEASET test 31504. [PWR

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

Booker, C.P.

TRAC-PD2/MOD1 is a publicly released version of TRAC that is used primarily to analyze large-break loss-of-coolant accidents in pressurized-water reactors (PWRs). TRAC-PD2 can calculate, among other things, reflood phenomena. TRAC posttest predictions are compared with test 31504 reflood data from the Full-Length Emergency Core Heat Transfer (FLECHT) System Effects and Separate Effects Tests (SEASET) facility. A false top-down quench is predicted near the top of the core and the subcooling is underpredicted at the bottom of the core. However, the overall TRAC predictions are good, especially near the center of the core.

Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

DOEpatents

DeVolpi, A.

1984-07-20

The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

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

Kim, Kang Seog; Clarno, Kevin T.; Gentry, Cole

2017-03-01

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

Instability study for LOFT for L2-1, L2-2, and L2-3 pretest steady-state operating conditions

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

Eide, S.A.

The results are presented of a thermal-hydrodynamic flow instability study of the LOFT reactor for the L2-1, L2-2, and L2-3 pretest steady-state operating conditions. Comparison is made between the LOFT reactor and a typical PWR, and the effects on stability of differences in operating parameters and geometry are discussed. Results indicate that the LOFT reactor will be thermal-hydrodynamically stable for nominal and worst case operating conditions. The study supports the LOFT Experimental Safety Analyses for the L2-1, L2-2, and L2-3 tests.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Efficient solution of the simplified P N equations

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.

2014-12-23

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

Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

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

Le Pape, Yann; Huang, Hai

Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

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.

Current and anticipated uses of thermal-hydraulic codes in Germany

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

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

1997-07-01

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

Design and implementation of a simple nuclear power plant simulator

NASA Astrophysics Data System (ADS)

Miller, William H.

1983-02-01

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

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

Rubber particle proteins, HbREF and HbSRPP, show different interactions with model membranes.

PubMed

Berthelot, Karine; Lecomte, Sophie; Estevez, Yannick; Zhendre, Vanessa; Henry, Sarah; Thévenot, Julie; Dufourc, Erick J; Alves, Isabel D; Peruch, Frédéric

2014-01-01

The biomembrane surrounding rubber particles from the hevea latex is well known for its content of numerous allergen proteins. HbREF (Hevb1) and HbSRPP (Hevb3) are major components, linked on rubber particles, and they have been shown to be involved in rubber synthesis or quality (mass regulation), but their exact function is still to be determined. In this study we highlighted the different modes of interactions of both recombinant proteins with various membrane models (lipid monolayers, liposomes or supported bilayers, and multilamellar vesicles) to mimic the latex particle membrane. We combined various biophysical methods (polarization-modulation-infrared reflection-adsorption spectroscopy (PM-IRRAS)/ellipsometry, attenuated-total reflectance Fourier-transform infrared (ATR-FTIR), solid-state nuclear magnetic resonance (NMR), plasmon waveguide resonance (PWR), fluorescence spectroscopy) to elucidate their interactions. Small rubber particle protein (SRPP) shows less affinity than rubber elongation factor (REF) for the membranes but displays a kind of "covering" effect on the lipid headgroups without disturbing the membrane integrity. Its structure is conserved in the presence of lipids. Contrarily, REF demonstrates higher membrane affinity with changes in its aggregation properties, the amyloid nature of REF, which we previously reported, is not favored in the presence of lipids. REF binds and inserts into membranes. The membrane integrity is highly perturbed, and we suspect that REF is even able to remove lipids from the membrane leading to the formation of mixed micelles. These two homologous proteins show affinity to all membrane models tested but neatly differ in their interacting features. This could imply differential roles on the surface of rubber particles. © 2013.