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Sample records for reactor pwr samples

  1. Library of PWR (pressurized-water reactor) steam generator tubing samples: Final report

    SciTech Connect

    Albertin, L.; Clark, W.G. Jr.; Junker, W.R.; Kuchirka, P.J.; Madeyski, A.; Metala, M.J.; Taszarek, B.J.

    1988-01-01

    The PWR Steam Generator Tubing Sample Library is a Steam Generator Owners Group-EPRI program whose objective is to compile a library of well-characterized tubing samples to be used for performance evaluation of inspection systems and for training and qualification of signal interpretation systems. The library was created through the preparation of samples intended to replicate degradation encountered in actual field tubes. A limited number of tube segments removed from actual steam generators are included. Degradation categories include wear, pitting and fatigue cracks, as well as stress corrosion cracking (SCC) and intergranular attack (IGA). Eddy current and ultrasonic inspection techniques, along with supplementary radiography, dye penetrant, and optical techniques were used to characterize the library candidates. Advanced computer-aided NDE data collection, analysis and display techniques were used to assess test results. This report provides details of the library program, with major emphasis on the sampling protocol, characterization of degradation and recommendations for the use and future growth of the library. Also included is a compendium of steam generator tube degradation field observation, describing past destructive examinations of tubes removed for inspection from steam generators, and a description of a physical modeling approach, using mercury (metal) to assess the discontinuity characterization capabilities of a pancake-type eddy current probe. Computerized data analysis and display techniques were used to reconstruct the test results in both two-dimensional color-coded maps and three-dimensional pseudo-isometric plots.

  2. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    SciTech Connect

    M. Krug; R. Shogan; A. Fero; M. Snyder

    2004-11-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR.

  3. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    SciTech Connect

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  4. Materials Reliability Program: Fracture Toughness Testing of Decommissioned PWR Core Internals Material Samples (MRP-160) Non-Proprietary Version

    SciTech Connect

    M. E. Krug; R. P. Shogan

    2005-09-30

    Pressurised water reactor (PWR) cores operate under extreme envrionmental conditions due to coolant chemistry, operating temperature and neutron exposure. Extending the life of PWRs requires detailed knowledge of teh changes in mechanical and corrosion properties of teh structural austenitic stainless steel components adjacent to the fuel. This report contains results of fracture toughness testing of samples machined from decommissioned PWR reactor internals.

  5. PWR systems transient analysis: a reactor-safety perspective

    SciTech Connect

    Kennedy, M.F.; Abramson, P.B.; McDonald, T.A.

    1982-01-01

    In the simulation of transient events in large PWR reactor systems for reactor safety studies, the plant model is quite detailed and must include most of the plant components and control systems to adequately analyze the range of transients. The results discussed were calculated with the RELAP4/MOD6 code and reveal the need for the analysis to carefully review and understand the results to assure that they are not being adversely affected by the improper solution techniques or changes in models during the calculation.

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

    SciTech Connect

    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 of the individual components equaling the measured values.

  7. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    SciTech Connect

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-07-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  8. Support arrangements for core modules of nuclear reactors. [PWR

    DOEpatents

    Bollinger, L.R.

    1983-11-03

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  9. 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.

  10. Nodal analysis for reactor kinetics and stability. [PWR; BWR

    SciTech Connect

    Park, J.K.; Becker, M.; Park, G.C.

    1983-07-01

    General space kinetics models have been developed for more accurate stability analysis utilizing nodal analysis, a commonly used technique for analyzing power distributions in large power reactors. Kinetics parameters for use in these kinetics models have been properly derived by utilizing self-consistent nodal data and power distributions. The procedure employed in the nodal code SIMULATE has been utilized for power distribution, since that methodology is general and includes various commonly used nodal methods as special cases. Cross sections are correlated as functions of void fraction and exposure. A computer program investigating thermo-hydrodynamic stability, NUFREQ has been modified to accommodate general spatial kinetics models with an improved thermal-hydraulics model. Stability analyses have been performed for density wave oscillations for a representative operating BWR system. Spatial coupling effects on the stability margins were found to be significant.

  11. Development of inspection systems for alloy 600 nozzles of PWR reactor vessel

    SciTech Connect

    Unate, K.; Ideo, M.; Sanagawa, T.; Shirai, T.; Araki, Y.

    1995-08-01

    PWR reactor vessels have alloy 600 nozzles at top and bottom heads. The former are head penetration nozzles for CRDM, and the latter are bottom mounted instrumentation nozzles. The authors have developed inspection systems of two types for each nozzle to confirm the soundness. ECT and UT Techniques are employed for both systems. These systems are controlled remotely and enable to reduce radiation exposure, inspection time and number of inspectors. Based on the functional tests using full scale mockups, the reliabilities and effectiveness of both systems were confirmed.

  12. Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients. [PWR

    SciTech Connect

    Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

    1980-01-01

    The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations.

  13. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  14. Bi-content Gadolinia as Burnable Absorber in PWR to Improve the Reactor Core Behaviour

    SciTech Connect

    Zheng, S.

    2007-07-01

    The gadolinia product is one of the standard burnable absorbers used in the PWR long and low leakage fuel cycle in order to control the radial power distribution and to hold down the initial core reactivity. This product presents a large number of advantages such as the high efficiency with only a small number of gadolinia-bearing rods, the easy adjustment between the number and the content of the gadolinia-bearing rods according to the cycle length need and the initial reactivity hold-down, no increasing of boron concentration versus cycle depletion, no additional increasing of internal pressure in poisoned rods, very low additional manufacture cost. On the other hand, some unfavourable phenomena are also observed during the utilization of the gadolinia: amplification of the asymmetrical power distribution and more negative axial offset. Based on the correlation between the gadolinia burnout and its content, the use of gadolinia bi-content will improve the parameters indicated here above. The gadolinia bi-content have been used in BWR for more than 20 years. In this paper, the comparison of the main reactor core physical parameters in PWR, calculated with the AREVA NP standard neutronic code package SCIENCE, is made by using the mono- and bi-content of the gadolinia products in the same fuel assembly. The results show that the asymmetrical axial and azimuthal power distribution can be improved in the case of the bi-content gadolinia product. (authors)

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

    DOE PAGESBeta

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; Wang, Jy-An John

    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

  16. VISA: a computer code for predicting the probability of reactor pressure-vessel failure. [PWR

    SciTech Connect

    Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.; Klecker, R.W.; Engel, D.W.; Johnson, K.I.

    1983-09-01

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

  17. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    SciTech Connect

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1985-04-01

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

  18. Sample problem calculations related to two-phase flow transients in a PWR relief-piping network

    SciTech Connect

    Shin, Y.W.; Wiedermann, A.H.

    1981-03-01

    Two sample problems related with the fast transients of water/steam flow in the relief line of a PWR pressurizer were calculated with a network-flow analysis computer code STAC (System Transient-Flow Analysis Code). The sample problems were supplied by EPRI and are designed to test computer codes or computational methods to determine whether they have the basic capability to handle the important flow features present in a typical relief line of a PWR pressurizer. It was found necessary to implement into the STAC code a number of additional boundary conditions in order to calculate the sample problems. This includes the dynamics of the fluid interface that is treated as a moving boundary. This report describes the methodologies adopted for handling the newly implemented boundary conditions and the computational results of the two sample problems. In order to demonstrate the accuracies achieved in the STAC code results, analytical solutions are also obtained and used as a basis for comparison.

  19. Analysis of the performance of the Westinghouse reactor vessel level indicating system for tests at semiscale. [PWR

    SciTech Connect

    Hardy, J.E.; Miller, G.N.

    1982-10-01

    The Westinghouse Reactor Vessel Level Indicating System (RVLIS), a differential pressure level measurement system, was tested at SEMISCALE. This report contains the analyses of these tests and the conclusions of these analyses. The tests performed included small break and intermediate break tests. Also, frequency response and natural circulation tests were run and analyzed. The RVLIS always indicated a level less than the two phase froth level. The RVLIS output in early small break tests indicated a level 200 cm greater than actual collapsed liquid level. This discrepancy was caused by structural differences between SEMISCALE and a Westinghouse reactor. Once modifications were made so that SEMISCALE better simulated a Westinghouse PWR, the maximum difference between RVLIS and SEMISCALE instrumentation was 30 cm or 3% which is less than the stated uncertainty of the Westinghouse RVLIS.

  20. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    SciTech Connect

    Shin, Y.W.; Wiedermann, A.H.

    1984-02-01

    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients.

  1. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  2. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, Louis K.; Alper, Naum I.

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  3. Probability of pipe failure in the reactor coolant loops of Babcock and Wilcox PWR plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1986-05-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) of reactor coolant piping as a design basis event for nuclear power plants, the US Nuclear Regulatory Commission (NRC) contracted the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence of a DEGB, and to assess the effect that earthquakes have on DEGB probability. This report describes an evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Babcock and Wilcox. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (''direct'' DEGB), and pipe rupture indirectly caused by failure of heavy component supports due to an earthquake (''indirect'' DEGB). Unlike in earlier evaluations of Westinghouse and Combustion Engineering reactor coolant loop piping, in which the probability of direct DEGB had been explicitly estimated using a probabilistic fracture mechanics model, no detailed fracture mechanics calculations were performed. Instead, a comparison of relevant plant data, mainly reactor coolant loop stresses, for one representative B and W plant with equivalent information for Westinghouse and C-E systems inferred that the probability of direct DEGB should be similarly low (less than le-10 per reactor year). The probability of indirect DEGB, on the other hand, was explicitly estimated for two representative plants. The results of this study indicate that the probability of a DEGB form either cause is very low for reactor coolant loop piping in these specific plants and, because of similarity in design, infer that the probability of DEGB is generally very low in B and W reactor coolant loop piping. The NRC should therefore consider eliminating DEGB as a design basis event in favor of more realistic criteria. 13 refs., 9 tabs.

  4. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Lo, T.; Chou, C.K.

    1985-01-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) as a design requirement for reactor coolant piping, the US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence of a DEGB, and to assess the effect that earthquakes have on DEGB probability. This report describes a probabilistic evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Combustion Engineering. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by failure of component supports due to an earthquake (indirect DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by estimating support fragility and then convolving fragility with seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for reactor coolant loop piping in these plants, and that NRC should therefore consider eliminating DEGB as a design basis in favor of more realistic criteria.

  5. Probability of pipe failure in the reactor coolant loops of Westinghouse PWR Plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1985-07-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) of reactor coolant loop piping as a design basis event for nuclear power plants, the US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence probability. This report describes a probabilistic evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Westinghouse. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (''direct'' DEGB), and pipe rupture indirectly caused by failure of component supports due to an earthquake (''indirect'' DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by estimating support fragility and then convolving fragility and seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for reactor coolant loop piping in these plants, and that NRC should therefore consider eliminating DEGB as a design basis event in favor of more realistic criteria. 17 refs., 15 figs., 11 tabs.

  6. 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.

  7. Dosimetry Evaluation of In-Core and Above-Core Zirconium Alloy Samples in a PWR

    NASA Astrophysics Data System (ADS)

    Amiri, Benjamin W.; Foster, John P.; Greenwood, Larry R.

    2016-02-01

    A description of the neutron fluence analysis of activated zirconium alloys samples at a Westinghouse 3-loop reactor is presented. These samples were irradiated in the core and in the fuel plenum region, where dosimetry measurements are relatively rare compared with regions radially outward of the core. Dosimetry measurements performed by Batelle/PNNL are compared to the calculational models. Good agreement is shown with the in-core measurements when using analysis conditions expected to best represent this region, such as an assembly-specific axial power distribution. However, the use of these conditions to evaluate dosimetry in the fuel plenum region can lead to significant underestimation of the fluence. The use of a flat axial power distribution, however, does not underestimate the fluence in the fuel plenum region.

  8. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  9. Reactor coolant pump startup under degraded conditions in a scaled OTSG lowered loop PWR

    SciTech Connect

    Tafreshi, A.M.; Marzo, M. di

    1996-12-31

    After a SB-LOCA or improper maintenance activities, the potential exists for a non-uniform distribution of boric acid in a PWR coolant system. This in turn presents the possibility of a reactivity excursion if sufficient volumes of boron-dilute water are transported into the core region without having first undergone substantial mixing. A research program is being conducted at the University of Maryland College Park (UMCP) 2 x 4 thermal-hydraulic test facility to assess the generation, transport and mixing of boron-dilute volumes. Start up of a pump and flow of a boron free slug of water in the cold leg and subsequent transport to the core downcomer in the facility is investigated here.

  10. Failure probability of PWR reactor coolant loop piping. [Double-ended guillotine break

    SciTech Connect

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria.

  11. Flow-induced vibration and instability of some nuclear-reactor-system components. [PWR

    SciTech Connect

    Chen, S.S.

    1983-01-01

    The high-velocity coolant flowing through a reactor system component is a source of energy that can induce component vibration and instability. In fact, many reactor components have suffered from excessive vibration and/or dynamic instability. The potential for detrimental flow-induced vibration makes it necessary that design engineers give detailed considerations to the flow-induced vibration problems. Flow-induced-vibration studies have been performed in many countries. Significant progress has been made in understanding the different phenomena and development of design guidelines to avoid damaging vibration. The purpose of this paper is to present an overview of the recent progress in several selected areas, to discuss some new results and to indentify future research needs. Specifically, the following areas will be presented: examples of flow-induced-vibration problems in reactor components; excitation mechanisms and component response characteristics; instability mechanisms and stability criteria; design considerations; and future research needs.

  12. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    SciTech Connect

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  13. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    SciTech Connect

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J.; Seren, T.; Lipponen, M.; Kekki, T.

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  14. Gamma-thermometer-based reactor-core liquid-level detector. [PWR

    SciTech Connect

    Burns, T.J.

    1981-06-16

    A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

  15. 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.

  16. 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

  17. Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors. [PWR; BWR

    SciTech Connect

    Chambers, R.; Laats, E.T.

    1981-01-01

    A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power history, fuel and cladding behavior models, and materials properties correlations. Three of the nine EMs provide either input or model specifications, or set the reference temperature for stored energy calculations. The remaining six EMs were intended to add thermal conservatism through model changes. To determine the relative influence of these six EMs upon fuel behavior calculations for commercial power reactors, a sensitivity study was conducted. That study is the subject of this paper.

  18. Halden In-Reactor Test to Exhibit PWR Axial Offset Anomaly

    SciTech Connect

    P.Bennett, B. Beverskog, R.Suther

    2004-12-01

    Many PWRs have encountered the axial offset anomaly (AOA) since the early 1990s, and these experiences have been reported widely. AOA is a phenomenon associated with localized boron hideout in corrosion product deposits (crud) on fuel surfaces. Several mitigation approaches have been developed or are underway to either delay the onset of AOA or avoid it entirely. This study describes the first phase of an experimental program designed to investigate whether the use of enriched boric acid (EBA) in the reactor coolant can mitigate AOA.

  19. Method for coupled three-dimensional analysis of reactor vessel blowdowns with internal structures. [PWR

    SciTech Connect

    Silling, S.A.; Gross, M.B.; Santee, G.E. Jr.; Chang, F.H.

    1981-01-01

    The STEALTH 3D and WHAMSE 3D computer codes have been combined to perform three-dimensional coupled fluid/structure calculations of the blowdown response of a pressure vessel with internal structures typical of a pressurized water reactor. The fluid/structure coupling, which is performed cycle by cycle during a calculation, is described. The coupled fluid/structure code, STEALTH/WHAMSE 3D, has been used to simulate the decompression of test V31.1 from the HDR blowdown test series. Calculations of fluid pressure, differential fluid pressure and hoop strain compare favorably with the experimental data from test V31.1. The computed peak axial stain compares less favorably with the experimental data, probably due to coarseness of the structural grid. 14 refs.

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

    NASA Technical Reports Server (NTRS)

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

    1986-01-01

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

  1. Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

    NASA Astrophysics Data System (ADS)

    Salko, Robert K.

    COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for

  2. Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

    NASA Astrophysics Data System (ADS)

    Salko, Robert K.

    COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for

  3. Retrospective Reactor Dosimetry with Zirconium Alloy Samples

    SciTech Connect

    Greenwood, Lawrence R.; Foster, John P.

    2009-11-01

    Retrospective measurements are routinely performed with stainless steel samples. Recent experiments have been successfully conducted using zirconium alloy samples, involving somewhat different neuron activation reactions than are normally encountered with stainless steel samples. The alloy composition consisted of nominally 1% (by weight) niobium, 1% tin, and 0.1% iron, with the balance zirconium. The activation products observed in the samples by gamma spectroscopy included Zr-95, Nb-95, Sn-113, Sb-125, Mn-54, Co-60, Nb-94, and Ta-182. The niobium was then chemically separated following ASTM procedure E1297 and the Nb-93m activities were measured by x-ray spectroscopy. The thermal neutron fluences, as determined independently by the neutron capture gamma reactions to Zr-95, Sn-113, Nb-94, and Sn/Sb-125, were in excellent agreement. The fast neutron fluences, as determined separately by the Fe-54(n,p)Mn-54 and Nb-93(n,n’)Nb-93m reactions, were also in good agreement, thus demonstrating the versatility of the retrospective dosimetry technique.

  4. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

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

    PubMed

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

    2009-09-01

    In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g. PMID:19297180

  6. Retrospective dosimetry analyses of reactor vessel cladding samples

    SciTech Connect

    Greenwood, L. R.; Soderquist, C. Z.; Fero, A. H.

    2011-07-01

    Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

  7. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail.

  8. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report. [PWR; BWR

    SciTech Connect

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1983-05-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces.

  9. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual. [PWR; BWR

    SciTech Connect

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code.

  10. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  11. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  12. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  13. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    SciTech Connect

    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.

  14. PWR fuel behavior: lessons learned from LOFT. [PWR

    SciTech Connect

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior.

  15. A Multi-Phased Sampling Effort to Characterize a University TRIGA Research Reactor

    SciTech Connect

    Taylor, K.E.; Holm, R.L.

    2006-07-01

    A radiological characterization project was conducted at the University of Illinois (University) TRIGA research nuclear reactor in July 2005 as part of the long-term facility decommissioning project. The characterization effort included multiple survey and sampling techniques designed to assess both contamination of the reactor building and equipment and activation of reactor components and the reactor bio-shield. Radiation measurements included alpha and beta surface contamination measurements, gamma dose rate measurements, and gross gamma radiation measurements. Modeling was conducted based on the field measurements to predict concentrations of activation products in reactor components that were not directly sampled. The sampling effort included collecting removable contamination swipes, concrete samples from the reactor room floor and bio-shield, soil samples from below and around the perimeter of the reactor building, graphite samples from graphite moderator, and metal samples from reactor components. Concrete samples were obtained using an innovative technology that allowed for quick sample collection and analysis. Concrete, soil, graphite, and metal samples were analyzed on-site using liquid scintillation counters and gamma spectroscopy. Additional samples were sent off-site for analysis. (authors)

  16. Sample Heat, Activity, Reactivity, and Dose Analysis for Safety Analysis of Irradiations in a Research Reactor.

    Energy Science and Technology Software Center (ESTSC)

    1987-12-01

    SHARDA is a program for assessing sample heating rates, activities produced and reactivity load caused while irradiating a small sample in a well thermalized research reactor like CIRUS. It estimates the sample cooling or lead shielding requirements to limit the gamma-ray dose rates due to the irradiated sample within permissible levels.

  17. Comparison of Removed Fuel Compositions of CANDLE, PWR, and FBR

    SciTech Connect

    Nagata, Akito; Sekimoto, Hiroshi

    2007-07-01

    A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replaced fresh fuels. About 40% of natural or depleted uranium undergoes fission. In this paper, spent fuels of PWR, FBR and CANDLE reactor are compared. Fresh fuels of PWR, FBR and CANDLE reactor are 4.1% enriched uranium (UO{sub 2}), MOX with 18.5% plutonium enrichment and natural uranium nitride (natural-UN), respectively. In once-through fuel cycle point of view, low disposal amount for high energy is better and CANDLE reactor can decrease this amount more than other reactors, especially it is only one-tenth of PWR fuel. Also, it can decrease MA and this amount is 0.4 times of PWR. Total FP amount of each reactor is nearly same. However, LLFP amount of CANDLE reactor is the largest. (authors)

  18. Estimating pressurized water reactor decommissioning costs: A user`s manual for the PWR Cost Estimating Computer Program (CECP) software. Draft report for comment

    SciTech Connect

    Bierschbach, M.C.; Mencinsky, G.J.

    1993-10-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  19. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  20. Neural net controlled tag gas sampling system for nuclear reactors

    DOEpatents

    Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

    1997-02-11

    A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

  1. Neural net controlled tag gas sampling system for nuclear reactors

    DOEpatents

    Gross, Kenneth C.; Laug, Matthew T.; Lambert, John D. B.; Herzog, James P.

    1997-01-01

    A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

  2. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    SciTech Connect

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  3. Ion chromatographic determination of transition metals in irradiated nuclear reactor surveillance samples.

    PubMed

    Louw, I

    1996-02-01

    The determination of transition metal ions in radioactive (+/-25 microCi/g) low-alloy steels (nuclear reactor surveillance samples) by ion chromatography (IC) is described. The analysis has been done directly without prior separation of the iron matrix. The eluted metal ions have been detected with a UV-visible spectrophotometric detector after post-column complexation with 4-(2-pyridylazo)resorcinol. The results are in a good agreement with the certified values for the standard reference material used. The method was applied to nuclear reactor surveillance samples for the determination of Cu, Mn, Co and Ni. PMID:15048428

  4. Defect formation in aqueous environment: Theoretical assessment of boron incorporation in nickel ferrite under conditions of an operating pressurized-water nuclear reactor (PWR)

    NASA Astrophysics Data System (ADS)

    Rák, Zs.; Bucholz, E. W.; Brenner, D. W.

    2015-06-01

    A serious concern in the safety and economy of a pressurized water nuclear reactor is related to the accumulation of boron inside the metal oxide (mostly NiFe2O4 spinel) deposits on the upper regions of the fuel rods. Boron, being a potent neutron absorber, can alter the neutron flux causing anomalous shifts and fluctuations in the power output of the reactor core. This phenomenon reduces the operational flexibility of the plant and may force the down-rating of the reactor. In this work an innovative approach is used to combine first-principles calculations with thermodynamic data to evaluate the possibility of B incorporation into the crystal structure of NiFe2O4 , under conditions typical to operating nuclear pressurized water nuclear reactors. Analyses of temperature and pH dependence of the defect formation energies indicate that B can accumulate in NiFe2O4 as an interstitial impurity and may therefore be a major contributor to the anomalous axial power shift observed in nuclear reactors. This computational approach is quite general and applicable to a large variety of solids in equilibrium with aqueous solutions.

  5. Leak before break application in French PWR plants under operation

    SciTech Connect

    Faidy, C.

    1997-04-01

    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.

  6. Sensitivity of risk parameters to human errors for a PWR

    SciTech Connect

    Samanta, P.; Hall, R. E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study.

  7. Flow through in situ reactors with suction lysimeter sampling capability and methods of using

    SciTech Connect

    Radtke, Corey W; Blackwelder, D Brad; Hubbell, Joel M

    2009-11-17

    An in situ reactor for use in a geological strata includes a liner defining a centrally disposed passageway and a sampling conduit received within the passageway. The sampling conduit may be used to receive a geological speciment derived from geological strata therein and a lysimeter is disposed within the sampling conduit in communication with the geological specimen. Fluid may be added to the geological specimen through the passageway defined by the liner, between an inside surface of the liner and an outside surface of the sampling conduit. A distal portion of the sampling conduit may be in fluid communication with the passageway.

  8. Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR

    SciTech Connect

    Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

    1983-08-01

    During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

  9. Influence Of Low Boron Core Design On PWR Transient Behavior

    SciTech Connect

    Aleksandrov Papukchiev, Angel; Yubo Liu; Schaefer, Anselm

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, the concentration of boron in primary coolant is limited by the requirement of having a negative moderator density coefficient. As high boron concentrations have significant impact on reactivity feedback properties, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) content has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) by approx. 50% compared to current German PWR technology. For the assessment of the potential safety advantages, a Loss-of-Feedwater Anticipated Transient Without Scram (ATWS LOFW) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The most significant difference in the transient performance of both designs is the total primary fluid mass released through the pressurizer (PRZ) valves. It is reduced by a factor of four for the low boron reactor, indicating its improved density reactivity feedback. (authors)

  10. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  11. Horizontal Drop of 21- PWR Waste Package

    SciTech Connect

    A.K. Scheider

    2007-01-31

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

  12. Fission-product data analysis from actinide samples exposed in the Dounreay Prototype Fast Reactor

    SciTech Connect

    Murphy, B.D.; Dickens, J.K.; Walker, R.L.; Newton, T.D.

    1994-12-31

    Since 1979 a cooperative agreement has been in effect between the United States and the United Kingdom to investigate the irradiation of various actinide species placed in the core of the Dounreay Prototype Fast Reactor (PFR). The irradiated species were isotopes of thorium, protactinium, uranium, neptunium, plutonium, americium, and curium. A set of actinide samples (mg quantities) was exposed to about 490 effective full power days (EFPD) of reactor operations. The fission-product results are reported here. The actinide results will be report elsewhere.

  13. Comparison of uranium determination in some Syrian geologic samples using three reactor based methods

    PubMed

    Jubeli

    2000-04-01

    A set of 25 samples of soil, sediments, carbonate and phosphate rocks from Syria were analysed for uranium, using three reactor based methods; instrumental neutron activation analysis (INAA), delayed neutron counting (DNC) and one cycle of irradiation utilizing the cyclic activation system (CAS). Although the three methods are capable of irradiation samples, the last method is the least established for U determination in rocks. The measurements obtained by the three methods are compared. The results show good agreement, with a distinct linear relationship and significant positive correlation coefficients. It was concluded that the CAS method could reliably be used to rapidly determine uranium in geological samples. PMID:10800739

  14. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR Plants. Volume 3. Double-ended guillotine break indirectly induced by earthquakes

    SciTech Connect

    Ravindra, M.K.; Campbell, R.D.; Kennedy, R.P.; Banon, H.

    1985-01-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of reactor coolant loop (RCL) piping have led to excessive design costs, interference of normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National Laboratory sponsored research program aimed at investigating whether the probability of DEGB in RCL Piping of nuclear power plants is acceptably small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimated the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of Combustion Engineering supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEGB was estimated to be in the range of 10/sup -6/ per year for older plants, and less than 10/sup -8/ per year for modern plants; using very conservative assumptions, the 90% subjective probability value (confidence) of P/sub DEGB/ was found to be less than 5 x 10/sup -5/ per year for older plants and less than 3 x 10/sup -7/ per year for modern plants.

  15. Probability of pipe failure in the reactor coolant loops of Babcock and Wilcox PWR plants. Volume 2. Guillotine break indirectly induced by earthquakes

    SciTech Connect

    Ravindra, M.K.; Campbell, R.D.; Kipp, T.R.; Sues, R.H.

    1985-07-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant loop (RCL) piping have led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National Laboratory sponsored research program aimed at exploring whether the probability of DEGB in RCL Piping of nuclear power plants is acceptably small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimates the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of Babcock and Wilcox supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEGB was estimated to range between 6 x 10/sup -11/ and 1 x 10/sup -7/ per year. Using very conservative assumptions, the 90% subjective probability value (confidence) of P/sub DEGB/ was found to be less than 1 x 10/sup -5/ per year. 19 refs., 19 figs., 11 tabs.

  16. Probability of pipe failure in the reactor coolant loop of Westinghouse PWR plants. Volume 3. Guillotine break indirectly induced by earthquakes

    SciTech Connect

    Ravindra, M.K.; Campbell, R.D.; Kennedy, R.P.; Banon, H.

    1985-02-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of reactor coolant loop (RCL) piping have led to excessive design costs, interference of normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National Laboratory sponsored research program aimed at investigating whether the probability of DEGB in RCL piping of nuclear power plants is acceptably small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimated the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of Westinghouse supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEGB was estimated to be about 3x10/sup -6/ per year with a 10% to 90% subjective probability range approximately from 1x10/sup -7/ per year to 5x10/sup -5/ per year.

  17. Determination of Long-Lived Neutron Activation Products in Reactor Shielding Concrete Samples

    SciTech Connect

    Zagar, Tomaz; Ravnik, Matjaz

    2002-10-15

    The results of activation studies of TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete were irradiated in the reactor to simulate neutron activation in the shielding concrete. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides were measured in the samples with a high-purity germanium detector. The most active long-lived radioactive nuclides in the ordinary concrete samples were found to be {sup 60}Co and {sup 152}Eu. In the barytes concrete samples, the most active long-lived radioactive nuclides were {sup 60}Co, {sup 133}Ba, and {sup 152}Eu. Activation in the concrete was also calculated using the ORIGEN2 code and compared to experimental results. Simple radioactive nuclide generation and depletion calculation using one-group cross-section libraries provided together with the ORIGEN2 code did not give conservative results. Significant discrepancies were observed for some nuclides. For accurate long-lived radioactive nuclide generation in reactor shielding, material-specific cross-section libraries should be generated and verified by measurement.

  18. Probability of pipe failure in the reactor coolant loops of Westinghouse PWR plants. Volume 4. Pipe failure induced by crack growth in west coast plants

    SciTech Connect

    Chinn, D.J.; Holman, G.S.; Lo, T.Y.; Mensing, R.W.

    1985-07-01

    The US Nuclear Regulatory Commission contracted with the Lawrence Livermore National Laboratory to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in primary coolant piping warrants the current design requirements that safeguard against the effecs of such a break. This report assesses the reactor-coolant-loop piping system of west coast Westinghouse plants. The results indicate that directly induced DEGB is an unlikely event in the west coast Westinghouse plants. For the Trojan plant, leak is far more likely than a direct DEGB. Further, earthquakes have very little effect on the probabilities of leak and direct DEGB. At the Diablo Canyon plant, the increase in postulated seismic levels due to reevaluation of the site to account for the Hosgri Fault has caused directly induced DEGB failure probability to be dependent on earthquake occurrences. The resulting direct DEGB failure probability is still much lower than the indirect DEGB failure probability for Diablo Canyon.

  19. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR plants. Volume 2. Pipe failure induced by crack growth

    SciTech Connect

    Lo, T.Y.; Mensing, R.W.; Woo, H.H.; Holman, G.S.

    1984-09-01

    A study was conducted to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the primary coolant piping warrants the current design requirements that safeguard against the effect of DEGB. This report describes the results of an assessment of reactor coolant loop piping systems designed by Combustion Engineering, Inc. A probabilistic fracture mechanics approach was used to estimate the crack growth and to assess the crack stability in the piping throughout the lifetime of the plant. The results of the assessment indicate that the probability of occurrence of DEGB due to crack growth and instability is extremely small, which supports the argument that the postulation of DEGB in design should be eliminated and replaced with more reasonable criteria.

  20. Long-term environmental trends: Selection of sampling locations in a reactor-aquatic cooling system

    SciTech Connect

    Revsin, B.K.; Watson, J.E. Jr. )

    1993-02-01

    The study objective was to determine whether environmental radionuclide accumulations were occurring in an aquatic system with a 13-y history of supplying a power plant with reactor-cooling water as well as receiving plant discharge. The aquatic system consisted of the following: (1) a reactor-cooling lake; (2) a secondary lake approximately 8 km downstream; and (3) a small stream that interfaced with the two lakes. Gamma-emitting radionuclides were identified and quantified in samples of benthic sediments obtained from representative areas of the aquatic system. This study demonstrated that in a reactor-aquatic cooling system, the component of the aquatic system most likely to experience radionuclide accumulation will not necessarily be the reactor-cooling lake, but will be that component of the aquatic system whose benthic sediments contain the highest concentrations of organic matter. Further, it was shown that the quantity of oxidizable organic matter present in a sediment is a good predictor or marker for potential sites of radionuclide accumulation (i.e., 60Co and 137Cs).

  1. Fukushima Daiichi reactor source term attribution using cesium isotope ratios from contaminated environmental samples

    DOE PAGESBeta

    Snow, Mathew S.; Snyder, Darin C.; Delmore, James E.

    2016-01-18

    Source term attribution of environmental contamination following the Fukushima Daiichi Nuclear Power Plant (FDNPP) disaster is complicated by a large number of possible similar emission source terms (e.g. FDNPP reactor cores 1–3 and spent fuel ponds 1–4). Cesium isotopic analyses can be utilized to discriminate between environmental contamination from different FDNPP source terms and, if samples are sufficiently temporally resolved, potentially provide insights into the extent of reactor core damage at a given time. Rice, soil, mushroom, and soybean samples taken 100–250 km from the FDNPP site were dissolved using microwave digestion. Radiocesium was extracted and purified using two sequentialmore » ammonium molybdophosphate-polyacrylonitrile columns, following which 135Cs/137Cs isotope ratios were measured using thermal ionization mass spectrometry (TIMS). Results were compared with data reported previously from locations to the northwest of FDNPP and 30 km to the south of FDNPP. 135Cs/137Cs isotope ratios from samples 100–250 km to the southwest of the FDNPP site show a consistent value of 0.376 ± 0.008. 135Cs/137Cs versus 134Cs/137Cs correlation plots suggest that radiocesium to the southwest is derived from a mixture of FDNPP reactor cores 1, 2, and 3. Conclusions from the cesium isotopic data are in agreement with those derived independently based upon the event chronology combined with meteorological conditions at the time of the disaster. In conclusion, cesium isotopic analyses provide a powerful tool for source term discrimination of environmental radiocesium contamination at the FDNPP site. For higher precision source term attribution and forensic determination of the FDNPP core conditions based upon cesium, analyses of a larger number of samples from locations to the north and south of the FDNPP site (particularly time-resolved air filter samples) are needed. Published in 2016. This article is a U.S. Government work and is in the public domain

  2. Light-water reactors: preliminary safety and environmental information document. Volume I

    SciTech Connect

    Not Available

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle (Pu/ThO/sub 2/ Burner).

  3. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  4. Residual life assessment of major light water reactor components: Overview

    SciTech Connect

    Shah, V.N.; MacDonald, P.E.; Amar, A.S.; Bakr, M.H.; Beaudoin, B.F.; Buescher, B.J.; Conley, D.A.; Drahos, F.R.; Gardner, J.B.; Garner, R.W.; Kirkwood, B.J.; Meyer, L.C.; Server, W.L.; Shah, V.N.; Siegel, E.A.; Sinha, U.P.; Ware, A.G. )

    1989-11-01

    This report presents an assessment of the aging (time-dependent degradation) of selected major light water reactor components and structures. The stressors, possible degradation sites and mechanisms, potential failure modes, and current inservice inspection requirements are discussed for eleven major light water reactor components: reactor coolant pumps, pressurized water reactor (PWR) pressurizers, PWR pressurizer surge and spray lines, PWR reactor coolant system charging and safety injection nozzles, PWR feedwater lines, PWR control rod drive mechanisms and reactor internals, boiling water reactor (BWR) containments, BWR feedwater and main steam lines, BWR control rod drive mechanisms and reactor internals, electrical cables and connections, and emergency diesel generators. Unresolved technical issues related to understanding and managing the aging of these major components are identified. 575 refs., 148 figs., 96 tabs.

  5. Characterisation of radiation field for irradiation of biological samples at nuclear reactor-comparison of twin detector and recombination methods.

    PubMed

    Golnik, N; Gryziński, M A; Kowalska, M; Meronka, K; Tulik, P

    2014-10-01

    Central Laboratory for Radiological Protection is involved in achieving scientific project on biological dosimetry. The project includes irradiation of blood samples in radiation fields of nuclear reactor. A simple facility for irradiation of biological samples has been prepared at horizontal channel of the nuclear reactor MARIA in NCBJ in Poland. The radiation field, composed mainly of gamma radiation and thermal neutrons, has been characterised in terms of tissue kerma using twin-detector technique and recombination chambers. PMID:24366246

  6. A target station for plasma exposure of neutron irradiated fusion material samples to reactor relevant conditions

    NASA Astrophysics Data System (ADS)

    Rapp, Juergen; Giuliano, Dominic; Ellis, Ronald; Howard, Richard; Lore, Jeremy; Lumsdaine, Arnold; Lessard, Timothy; McGinnis, William; Meitner, Steven; Owen, Larry; Varma, Venugopal

    2015-11-01

    The Material Plasma Exposure eXperiment (MPEX) is a device planned to address scientific and technological gaps for the development of viable plasma facing components for fusion reactor conditions (FNSF, DEMO). It will have to address the relevant plasma conditions in a reactor divertor (electron density, electron temperature, ion fluxes) and it needs to be able to expose a-priori neutron irradiated samples. A pre design of a target station able to handle activated materials will be presented. This includes detailed MCNP as well as SCALE and MAVRIC calculations for all potential plasma-facing materials to estimate dose rates. Details on the remote handling schemes for the material samples will be presented. 2 point modeling of the linear plasma transport has been used to scope out the parameter range of the anticipated power fluxes to the target. This has been used to design the cooling capability of the target. The operational conditions of surface temperatures, plasma conditions, and oblique angle of incidence of magnetic field to target surface will be discussed. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under contract DE-AC-05-00OR22725.

  7. STEALTH: a Lagrange explicit finite difference code for solids, structural, and thermohydraulic analysis. Volume 2: sample and verification problems. Computer code manual. [PWR; BWR

    SciTech Connect

    Hofmann, R.

    1982-08-01

    STEALTH sample and verification problems are presented to help users become familiar with STEALTH capabilities, input, and output. Problems are grouped into articles which are completely self-contained. The pagination in each article is A.n, where A is a unique alphabetic-character article identifier and n is a sequential page number which starts from 1 on the first page of text for each article. Articles concerning new capabilities will be added as they become available. STEALTH sample and verification calculations are divided into the following general categories: transient mechanical calculations dealing with solids; transient mechanical calculations dealing with fluids; transient thermal calculations dealing with solids; transient thermal calculations dealing with fluids; static and quasi-static calculations; and complex boundary interaction calculations.

  8. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

    SciTech Connect

    Klann, R. T.; Perret, G.; Nuclear Engineering Division

    2007-10-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.

  9. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect

    Wang, S.-J.; Chiang, K.-S.; Chiang, S.-C.

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  10. Evaluation of thermal mixing data from a model cold leg and downcomer. [PWR

    SciTech Connect

    Rothe, P.H.; Fanning, M.W.

    1982-12-01

    This report describes an evaluation of thermal mixing data obtained in a 1/5-scale, transparent model of the cold leg and downcomer of a Pressurized Water Reactor (PWR). The data are relevant to the phenomenon of fluid and thermal mixing following HPI (High Pressure Injection) of coolant water in a PWR loop. The data are reduced, correlated and compared with theoretically derived values and scaling approaches.

  11. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    SciTech Connect

    Not Available

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

  12. Analysis of MERCI decay heat measurement for PWR UO{sub 2} fuel rod

    SciTech Connect

    Jaboulay, J.C.; Bourganel, S.

    2012-01-15

    Decay heat measurements, called the MERCI experiment, were conducted at Commissariat a l'Energie Atomique (CEA)/Saclay to characterize accurately residual power at short cooling time and verify its prediction by decay code and nuclear data. The MOSAIC calorimeter, developed and patented by CEA/Grenoble (DTN/SE2T), enables measurement of the decay heat released by a pressurized water reactor (PWR) fuel rod sample between 200 and 4 W within a precision of 1%. The MERCI experiment included three phases. At first, a UO{sub 2} fuel rod sample was irradiated in the CEA/Saclay experimental reactor OSIRIS. The burnup achieved at the end of irradiation was similar to 3.5 GWd/tonne. The second phase was the transfer of the fuel rod sample from its irradiation location to a hot cell, to be inserted inside the MOSAIC calorimeter. It took 26 min to carry out the transfer. Finally, decay heat released by the PWR sample was measured from 27 min to 42 days after shutdown. Post irradiation examinations were performed to measure concentrations of some heavy nuclei (U, Pu) and fission products (Cs, Nd). The decay heat was predicted using a calculation scheme based on the PEPIN2 depletion code, the TRIPOLI-4 Monte Carlo code, and the JEFF3.1.1 nuclear data file. The MERCI experiment analysis shows that the discrepancy between the calculated and the experimental decay heat values is included between -10% at 27 min and +6% at 12 h, 30 min otter shutdown. From 4 up to 42 days of cooling time, the difference between calculation and measurement is about ± 1%, i.e., experimental uncertainty. The MERCI experiment represents a significant contribution for code validation; the time range above 10{sup 5} s has not been validated previously. (authors)

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

    SciTech Connect

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    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 cases 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.

  14. iDREAM: an industrial detector for nuclear reactor monitoring

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

    Prototype of industrial reactor antineutrino detector iDREAM is dedicated for an experiment to demonstrate the possibility of remote monitoring of PWR reactor operational modes by neutrino method in real-time in order to avoid undeclared exposure modes for nuclear fuel and unauthorized removal of isotopes. The prototype detector was started up in 2014. To test the detector elements and components of electronics distilled water has been used as a target, which enables the use of Cerenkov radiation from cosmic muons as a physical signal. Also parallel measuring of the long-term stability has been doing for samples of liquid organic scintillator doped with gadolinium and synthesized by different methods

  15. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  16. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  17. Decontamination as a precursor to decommissioning. Status report Task 2: process evaluation. [PWR; BWR

    SciTech Connect

    Divine, J.R.; Woodruff, E.M.; McPartland, S.A.; Zima, G.E.

    1983-05-01

    As part of the US Nuclear Regulatory Commission's program to reduce occupational exposure and waste volumes, the Pacific Northwest Laboratory is studying decontamination as a precursor to decommissioning. Eleven processes or solvents were examined for their behavior in decontaminating BWR carbon steel samples. The solvents included NS-1, a proprietary solvent of Dow Chemical Corporation, designed for BWR use, and AP-Citrox, a well-known, two-step process designed for PWR stainless steel; it was used to provide a reference for later comparison to other systems and processes. The decontamination factors observed in the tests performed in a small laboratory scale recirculating loop ranged from about 1 (no effect) to 222 (about 99.6% of the initial activity removed. Coordinated corrosion measurements were made using twelve chemical solvents and eight metal alloys found in a range of reactor types.

  18. A Study on the Conceptual Design of a 1,500 MWe Passive PWR with Annular Fuel

    SciTech Connect

    Kwi Lim Lee; Soon Heung Chang

    2004-07-01

    In this study, the preliminary conceptual design of a 1500 MWe pressurized water reactor (PWR) with annular fuel has been performed. This design is derived from the AP1000 which is a 1000 MWe PWR with two-loop. However, the present design is a 1500 MWe PWR with three-loop, passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. The preliminary design parameters of this reactor have been determined through simple relation to those of AP1000 for reactor, reactor coolant system, and passive safety injection system. Using the MATRA code, we analyze the core designs for two alternatives on fuel assembly types: solid fuel and annular fuel. The performance of reactor cooling systems is evaluated through the accident of the cold leg break in the core makeup tank loop by using MARS2.1 code. This study presents the developmental strategy, preliminary design parameters and safety analysis results. (authors)

  19. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    SciTech Connect

    Kavaklioglu, K.; Ikonomopoulos, A. )

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint.

  20. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  1. Characterization of an aerosol sample from the auxiliary building of the Three Mile Island reactor.

    PubMed

    Kanapilly, G M; Stanley, J A; Newton, G J; Wong, B A; DeNee, P B

    1983-11-01

    Analyses for radioisotopic composition and dissolution characteristics were performed on an aerosol filter sample collected for a week by an air sampler located in the auxiliary building of the Three Mile Island nuclear reactor. The major radioisotopes found on the filter were 89Sr, 90Sr, 134Cs and 137Cs. Greater than 90% of both 89-90Sr and 134-137Cs dissolved within 48 hr in an in vitro test system. Scanning electron microscopic analyses showed the presence of respirable size particles as well as larger particles ranging up to 10 micron in diameter. The major matrix components were Fe, Ca, S, Mg, Al and Si. Although the radionuclides were present in a heterogeneous matrix, they were in a soluble form. This information enables a better evaluation of bioassay data and predictions of dose distribution resulting from an inhalation exposure to this aerosol. Further, the combination of techniques used in this study may be applicable to the characterization of other aerosols of unknown composition. PMID:6643066

  2. RIA Limits Based On Commercial PWR Core Response To RIA

    SciTech Connect

    Beard, Charles L.; Mitchell, David B.; Slagle, William H.

    2006-07-01

    Reactivity insertion accident (RIA) limits have been under intense review by regulators since 1993 with respect to what should be the proper limit as a function of burnup. Some national regulators have imposed new lower limits while in the United States the limits are still under review. The data being evaluated with respect to RIA limits come from specialized test reactors. However, the use of test reactor data needs to be balanced against the response of a commercial PWR core in setting reasonable limits to insure the health and safety of the public without unnecessary restrictions on core design and operation. The energy deposition limits for a RIA were set in the 1970's based on testing in CDC (SPERT), TREAT, PBF and NSRR test reactors. The US limits given in radially averaged enthalpy are 170 cal/gm for fuel cladding failure and 280 cal/gm for coolability. Testing conducted in the 1990's in the CABRI, NSRR and IGR test reactors have demonstrated that the cladding failure threshold is reduced with burnup, with the primary impact due to hydrogen pickup for in-reactor corrosion. Based on a review of this data very low enthalpy limits have been proposed. In reviewing proposed limits from RIL-0401(1) it was observed that much of the data used to anchor the low allowable energy deposition levels was from recent NSRR tests which do not represent commercial PWR reactor conditions. The particular characteristics of the NSRR test compared to commercial PWR reactor characteristics are: - Short pulse width: 4.5 ms vs > 8 ms; - Low temperature conditions: < 100 deg. F vs 532 deg. F. - Low pressure environment: atmospheric vs {approx} 2200 psi. A review of the historical RIA database indicates that some of the key NSRR data used to support the RIL was atypical compared to the overall RIA database. Based on this detailed review of the RIA database and the response of commercial PWR core, the following view points are proposed. - The Failure limit should reflect local fuel

  3. Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor

    NASA Astrophysics Data System (ADS)

    Ródenas, J.; Gallardo, S.; Weirich, F.; Hansen, W.

    2014-11-01

    All materials present in the core of a nuclear reactor are activated by neutron irradiation. The activity so generated produces a dose around the material. This dose is a potential risk for workers in the surrounding area when materials are withdrawn from the reactor. Therefore, it is necessary to assess the activity generated and the dose produced. In previous works, neutron activation of control rods and doses around the storage pool where they are placed have been calculated for a Boiling Water Reactor using the MCNP5 code based on the Monte Carlo method. Most of the activation is produced indeed in stainless steel components of the nuclear reactor core not only control rods. In this work, a stainless steel sample is irradiated in the Training Reactor AKR-2 of the Technical University Dresden. Dose measurements around the sample have been performed for different times after the irradiation. Experimental dosimetric values are compared with results of Monte Carlo simulation of the irradiation. Comparison shows a good agreement. Hence, the activation Monte Carlo model can be considered as validated.

  4. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  5. Simple automatic device for real time sampling of gas production by a reactor

    SciTech Connect

    Frattolillo, A.

    2006-06-15

    An innovative automatic device, allowing periodically drawing samples of the gases produced by a generic reactor, is presented. The gases evolving during the reaction are collected in a storage manifold, equipped with a variable volume consisting of a stainless steel bellow, whose expansion or contraction is driven by a linear step motor. A capacitive gauge monitors the pressure inside the storage manifold, while a feedback control loop reacts to any pressure change adjusting the variable volume (by means of the step motor) in such a way to keep the pressure at a desired set point P{sub 0}. As long as the reaction proceeds, the gas production results in a progressive expansion of the variable volume, whose instantaneous value is constantly monitored by means of a slide potentiometer, whose lever is rigidly connected to the bellow's moving extremity. Once the bellow's expansion has reached a predetermined volume increment {delta}V, which means that an amount of gas P{sub 0}{delta}V has been produced and collected in the storage circuit, a quantity P{sub 0}V{sub S}{<=}P{sub 0}{delta}V of gas is released to the analysis system. A set of electropneumatic valves, automatically operated by the control system, allows for gas delivery to the analysis equipment and retrieval of the set point pressure, by compression of the variable volume, with no influence on the reaction. All relevant parameters are monitored and logged on a personal computer. The control and data acquisition software, made out using National Instrument LABVIEW trade mark sign , also provides control of the analysis equipment. The ability of the proposed setup to not affect the ongoing process allows real time monitoring (by drawing samples at regular time intervals during the reaction) of the gas production. Moreover, since the amount of gas P{sub 0}V{sub S} drawn at each sampling is always the same, it is possible to establish at a glance whether or not there are changes in the concentration of any

  6. Enriched boric acid for PWR application: Cost evaluation study for a twin-unit PWR

    SciTech Connect

    Battaglia, J.A.; Waters, R.M.; von Hollen, J.M.; Lamatia, L.A.; Bergmann, C.A.; Ditommaso, S.M. . Nuclear and Advanced Technology Div.)

    1989-09-01

    In the nuclear industry boric acid dissolved in the reactor coolant is used as a soluble reactivity control agent. Reactivity control in nuclear plants is also provided by neutron absorbing control rods. This neutron absorbing duty is distributed between the control rods and soluble boric acid in such a way as to provide the most economical split. Typically, the control rods take care of rapid reactivity changes and the boric acid handles the slower long term control of reactivity by varying the boric acid concentrations within the reactor coolant. In PWR reactor plants the dissolved boric acid is referred to as a soluble poison or chemical shim due to the high capacity for thermal neutron capture exhibited by the boron-10 isotope contained in the boric acid molecule. This slow reactivity change or chemical shim control would otherwise have to be performed using control rods, a much more expensive proposition. Reactivity changes are controlled by the B-10 isotope by virtue of its very high cross section (3837 barns) for thermal neutron absorption. However, natural boron contains only 20 atom percent of the B-10 isotope and essentially all the remaining 80 percent as the B-11 isotope. The B-11 isotope of cross section .005 barns is essentially of no use as a neutron absorber. Since B-11 makes up the bulk of the total boron present and contributes little to the nuclear operation it would seem logical to eliminate this isotope of boron from the boric acid molecule. In so doing boric acid concentration in operating PWR plants need only be a fraction of that existing to accomplish identical nuclear operations. However, to achieve the elimination of B-11 from NBA (Natural Boric Acid) an isotope separation must be performed. 4 refs., 25 figs., 17 tabs.

  7. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    SciTech Connect

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-22

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

  8. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    NASA Astrophysics Data System (ADS)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-01

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required 233U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium & uranium confinement in PWR.

  9. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  10. High Cycle Thermal Fatigue in French PWR

    SciTech Connect

    Blondet, Eric; Faidy, Claude

    2002-07-01

    Different fatigue-related incidents which occurred in the world on the auxiliary lines of the reactor coolant system (SIS, RHR, CVC) have led EDF to search solutions in order to avoid or to limit consequences of thermodynamic phenomenal (Farley-Tihange, free convection loop and stratification, independent thermal cycling). Studies are performed on mock-up and compared with instrumentation on nuclear power stations. At the present time, studies allow EDF to carry out pipe modifications and to prepare specifications and recommendations for next generation of nuclear power plants. In 1998, a new phenomenal appeared on RHR system in Civaux. A crack was discovered in an area where hot and cold fluids (temperature difference of 140 deg. C) were mixed. Metallurgic studies concluded that this crack was caused by high cycle thermal fatigue. Since 1998, EDF is making an inventory of all mixing areas in French PWR on basis of criteria. For all identified areas, a method was developed to improve the first classifying and to keep back only potential damage pipes. Presently, studies are performing on the charging line nozzle connected to the reactor pressure vessel. In order to evaluate the load history, a mock-up has been developed and mechanical calculations are realised on this nozzle. The paper will make an overview of EDF conclusions on these different points: - dead legs and vortex in a no flow connected line; - stratification; - mixing tees with high {delta}T. (authors)

  11. CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    SciTech Connect

    Crawford, C.; Bannochie, C.

    2014-05-12

    A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD

  12. Fast Pyrolysis of Poplar Using a Captive Sample Reactor: Effects of Inorganic Salts on Primary Pyrolysis Products

    SciTech Connect

    Mukarakate, C.; Robichaud, D.; Donohoe, B.; Jarvis, M.; Mino, K.; Bahng, M. K.; Nimlos, M.

    2012-01-01

    We have constructed a captive sample reactor (CSR) to study fast pyrolysis of biomass. The reactor uses a stainless steel wire mesh to surround biomass materials with an isothermal environment by independent controlling of heating rates and pyrolysis temperatures. The vapors produced during pyrolysis are immediately entrained and transported in He carrier gas to a molecular beam mass spectrometer (MBMS). Formation of secondary products is minimized by rapidly quenching the sample support with liquid nitrogen. A range of alkali and alkaline earth metal (AAEM) and transition metal salts were tested to study their effect on composition of primary pyrolysis products. Multivariate curve resolution (MCR) analysis of the MBMS data shows that transition metal salts enhance pyrolysis of carbohydrates and AAEM salts enhances pyrolysis of lignin. This was supported by performing similar separate studies on cellulose, hemicellulose and extracted lignin. The effect of salts on char formation is also discussed.

  13. Design study of long-life PWR using thorium cycle

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

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

  14. Design study of long-life PWR using thorium cycle

    SciTech Connect

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

    2012-06-06

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

  15. PWR Cross Section Libraries for ORIGEN-ARP

    SciTech Connect

    McGraw, Carolyn; Ilas, Germina

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  16. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  17. VERA Core Simulator Methodology for PWR Cycle Depletion

    SciTech Connect

    Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel; Kim, Kang Seog; Graham, Aaron; Stimpson, Shane; Wieselquist, William A; Clarno, Kevin T; Palmtag, Scott; Downar, Thomas; Gehin, Jess C

    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 nuclear 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.

  18. Electropolishing process development for PWR steam generator channel heads

    SciTech Connect

    Asay, R.H.; Graves, P.; Guastaferro, C.T.; Spalaris, C.N. )

    1991-04-01

    A broad range of process parameters was established to smoothen the surface of 309 L weld clad overlay, prototypic of surfaces common is channel heads of replacement PWR (pressurized water reactor) steam generators. Mechanical and electropolishing steps were studied to explore process boundaries, which result in acceptable degree of surface smoothness, without compromising metallurgical properties. Recommended processes and acceptance criteria established in this work, can be applied to electropolish steam generator channel heads. Smooth surfaces are less likely to retain radioactive species, and potentially develop lower radiation fields when these components are placed into service. 7 refs., 11 figs., 12 tabs.

  19. Project of the new multifunctional reflectometer GRAINS with horizontal sample plane at the IBR-2M pulsed reactor in Dubna

    NASA Astrophysics Data System (ADS)

    Avdeev, M. V.; Bodnarchuk, V. I.; Lauter-Pasyuk, V. V.; Lauter, H.; Aksenov, V. L.; Yaradaikin, S. P.; Ulyanov, V. A.; Trounov, V. A.; Kalinin, S. I.

    2010-11-01

    The new multifunctional reflectometer GRAINS is under construction at the modernized high flux IBR-2 pulsed reactor (IBR-2M) of the FLNP JINR (Dubna, Russia). The principal feature of this reflectometer, the horizontal sample plane (or vertical scattering plane), enables the study of liquid-containing interfaces. The reflectometer will operate in the time-of-flight regime with constant sample illumination during measurements. The important advantage is the constant angular resolution with unvaried background. The additional modes of the GRAINS reflectometer comprise (1) off-specular scattering and GISANS, which are measured simultaneously in the TOF regime at a 2D position-sensitive detector; (2) angular encoding in the horizontal plane, which is provided by a Larmor precession region limited by current sheets in front of the sample; (3) 3D polarimetry in reflection, which is provided by a Larmor precession region around the sample position. The design of the reflectometer is optimised to take better advantage of an exceptionally broad wavelength band of the new cold moderator at the IBR-2M. The set-up will open up principally new possibilities for investigations in the field of interface nano-science at the IBR-2M reactor.

  20. ASSESSMENT OF THE RADIONUCLIDE COMPOSITION OF "HOT PARTICLES" SAMPLED IN THE CHERNOBYL NUCLEAR POWER PLANT FOURTH REACTOR UNIT

    SciTech Connect

    Farfan, E.; Jannik, T.; Marra, J.

    2011-10-01

    Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) 4th Reactor Unit Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified and the fuel burnup in these samples was determined. A systematic deviation in the burnup values based on the cesium isotopes, in comparison with other radionuclides, was observed. The conducted studies were the first ever performed to demonstrate the presence of significant quantities of {sup 242}Cm and {sup 243}Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from {sup 241}Am (and going higher), in comparison with the theoretical calculations.

  1. Turning points in reactor design

    SciTech Connect

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

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

    SciTech Connect

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

    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 an 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)

  3. Transport and reaction of nanoliter samples in a microfluidic reactor using electro-osmotic flow

    NASA Astrophysics Data System (ADS)

    Arumbuliyur Comandur, Kaushik; Bhagat, Ali Asgar S.; Dasgupta, Subhashish; Papautsky, Ian; Banerjee, Rupak K.

    2010-03-01

    The primary focus of the paper is to establish both numerical and experimental methods to control the concentration of samples in a microreactor well. The concentration of the reacting samples is controlled by varying the initial sample size and electric field. Further, the paper numerically investigates the feasibility of mixing and reacting nanoliter samples with a wide variation in reaction rates in the microreactor driven by electro-osmotic pumping. Two discrete samples are measured and transported to the microreactor simultaneously by electro-osmotic pinching and switching. The transported samples are mixed in the microreactor and floated for 4.5 s for reaction to occur. It is seen that the normalized concentration of the product increases from 0.25 to 0.45 during that period. Also the effects of sample size and applied electric field on sample concentration during the switching process are studied. It is found that the normalized final sample concentration increases from 0.03 to 0.11 with an increase in sample size from 60 to 150 µm, at a constant electric field. Further, by increasing the electric field from 100 to 1000 V cm-1, at a constant sample size, there is a significant decrease in the final concentration of the sample from 0.14 to 0.04. Our studies also show that the normalized product concentration depends on the reaction rate and increases from 0.28 to 0.48 as the reaction rate increases from 10 L mol-1 s-1 to 105 L mol-1 s-1. However, the increase in the reaction rate beyond 105 L mol-1 s-1 does not influence the product concentration for the present design of the microreactor. Our microreactor with improved mixing can be used for assessing reactions of biological samples. The optimized sample size along with a controlled electric field for sample injection forms the basis for developing a prototype of a microreactor device for high throughput drug screening.

  4. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    NASA Astrophysics Data System (ADS)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  5. Timing analysis of PWR fuel pin failures

    SciTech Connect

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. ); Straka, M. )

    1992-09-01

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively.

  6. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    Energy Science and Technology Software Center (ESTSC)

    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 maymore » be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 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

  7. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    SciTech Connect

    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 69 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

  8. Long-lived activation products in reactor materials

    SciTech Connect

    Evans, J.C.; Lepel, E.L.; Sanders, R.W.; Wilkerson, C.L.; Silker, W.; Thomas, C.W.; Abel, K.H.; Robertson, D.R.

    1984-08-01

    The purpose of this program was to assess the problems posed to reactor decommissioning by long-lived activation products in reactor construction materials. Samples of stainless steel, vessel steel, concrete, and concrete ingredients were analyzed for up to 52 elements in order to develop a data base of activatable major, minor, and trace elements. Large compositional variations were noted for some elements. Cobalt and niobium concentrations in stainless steel, for example, were found to vary by more than an order of magnitude. A thorough evaluation was made of all possible nuclear reactions that could lead to long lived activation products. It was concluded that all major activation products have been satisfactorily accounted for in decommissioning planning studies completed to date. A detailed series of calculations was carried out using average values of the measured compositions of the appropriate materials to predict the levels of activation products expected in reactor internals, vessel walls, and bioshield materials for PWR and BWR geometries. A comparison is made between calculated activation levels and regulatory guidelines for shallow land disposal according to 10 CFR 61. This analysis shows that PWR and BWR shroud material exceeds the Class C limits and is, therefore, generally unsuitable for near-surface disposal. The PWR core barrel material approaches the Class C limits. Most of the remaining massive components qualify as either Class A or B waste with the bioshield clearly Class A, even at the highest point of activation. Selected samples of activated steel and concrete were subjected to a limited radiochemical analysis program as a verification of the computer model. Reasonably good agreement with the calculations was obtained where comparison was possible. In particular, the presence of /sup 94/Nb in activated stainless steel at or somewhat above expected levels was confirmed.

  9. Assessment of void swelling in austenitic stainless steel PWR core internals.

    SciTech Connect

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and

  10. User's guide for the PWR LOCA analysis capability of the WRAP-EM system

    SciTech Connect

    Beranek, F; Gregory, M V

    1980-02-01

    The Water Reactor Analysis Package (WRAP) has been expanded to provide the capability to analyze loss-of-coolant accidents (LOCAs) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluation models (EMs). The input specifications for modules in the WRAP-EM system are presented in this document along with the JOSHUA input templates. This document, along with the WRAP user's guide, provides a step-by-step procedure for setting up a PWR data base for the WRAP-EM system. 12 refs.

  11. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    NASA Astrophysics Data System (ADS)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  12. Chelatococcus caeni sp. nov., isolated from a biofilm reactor sludge sample.

    PubMed

    Jin, Long; Ko, So-Ra; Lee, Hyung-Gwan; Kim, Byung-Hyuk; Kim, Hee-Sik; Ahn, Chi-Yong; Oh, Hee-Mock

    2015-03-01

    A polyphasic taxonomic study was carried out on strain EBR-4-1(T), which was isolated from a biofilm reactor in the Republic of Korea. The cells of the strain were Gram-stain-negative, non-spore-forming, motile and rod-shaped. Comparative 16S rRNA gene sequence studies showed a clear affiliation of this strain to the Alphaproteobacteria, and it was most closely related to Chelatococcus daeguensis CCUG 54519(T), Chelatococcus sambhunathii HT4(T), and Chelatococcus asaccharovorans DSM 6462(T) with 16S rRNA gene sequence similarities to the type strains of these species of 98.8 %, 98.7 %, and 96.3 %, respectively. The G+C content of the genomic DNA of strain EBR-4-1(T) was 68.7 mol%. Phenotypic and chemotaxonomic data [Q-10 as the major ubiquinone; C19 : 0cycloω8c, C18 : 1 2-OH, and summed feature 8 (C18 : 1ω7c and/or C18 : 1ω6c) as the major fatty acids] supported the affiliation of strain EBR-4-1(T) to the genus Chelatococcus. On the basis of the polyphasic evidence, it is proposed that strain EBR-4-1(T) should be assigned to a new species, Chelatococcus caeni sp. nov. The type strain is EBR-4-1(T) ( = KCTC 32487(T) = JCM 30181(T)). PMID:25525126

  13. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  14. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  15. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  16. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect

    J.M. Scaglione

    2004-12-17

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

  17. Radial diffusion and penetration of gas molecules and aerosol particles through laminar flow reactors, denuders, and sampling tubes.

    PubMed

    Knopf, Daniel A; Pöschl, Ulrich; Shiraiwa, Manabu

    2015-04-01

    Flow reactors, denuders, and sampling tubes are essential tools for many applications in analytical and physical chemistry and engineering. We derive a new method for determining radial diffusion effects and the penetration or transmission of gas molecules and aerosol particles through cylindrical tubes under laminar flow conditions using explicit analytical equations. In contrast to the traditional Brown method [Brown, R. L. J. Res. Natl. Bur. Stand. (U. S.) 1978, 83, 1-8] and CKD method (Cooney, D. O.; Kim, S. S.; Davis, E. J. Chem. Eng. Sci. 1974, 29, 1731-1738), the new approximation developed in this study (known as the KPS method) does not require interpolation or numerical techniques. The KPS method agrees well with the CKD method under all experimental conditions and also with the Brown method at low Sherwood numbers. At high Sherwood numbers corresponding to high uptake on the wall, flow entry effects become relevant and are considered in the KPS and CKD methods but not in the Brown method. The practical applicability of the KPS method is demonstrated by analysis of measurement data from experimental studies of rapid OH, intermediate NO3, and slow O3 uptake on various organic substrates. The KPS method also allows determination of the penetration of aerosol particles through a tube, using a single equation to cover both the limiting cases of high and low deposition described by Gormley and Kennedy (Proc. R. Ir. Acad., Sect. A. 1949, 52A, 163-169). We demonstrate that the treatment of gas and particle diffusion converges in the KPS method, thus facilitating prediction of diffusional loss and penetration of gases and particles, analysis of chemical kinetics data, and design of fluid reactors, denuders, and sampling lines. PMID:25744622

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

    SciTech Connect

    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 also been shown to work seamlessly with the Uniform Fission Site (UFS) variance reduction method. (authors)

  19. Material Sample Collection with Tritium and Gamma Analyses at the University of Illinois's Nuclear Research Laboratory TRIGA Nuclear Research Reactor

    SciTech Connect

    Charters, G.; Aggarwal, S.

    2006-07-01

    The University of Illinois in Champaign-Urbana has an Advanced TRIGA reactor facility which was built in 1960 and operated until August 1998. The facility was shutdown for a variety of reasons, primarily due to a lack of usage by the host institution. In 1998 the reactor went into SAFSTOR and finally shipped its fuel in 2004. At the present time a site characterization and decommissioning plan are in process and hope to be submitted to the NRC in early 2006. The facility had to be fully characterized and part of this characterization involved the collection and analysis of samples. This included various solid media such as, concrete, graphite, metals, and sub-slab surface soils for immediate analysis of Activation and Tritium contamination well below the easily measured surfaces. This detailed facility investigation provided a case to eliminate historical unknowns, increasing the confidence for the segregation and packaging of high specific activity Low Level Radwaste (LLRW), from which a strategy of 'surgical-demolition' and segregation could be derived thus maximizing the volumes of 'clean material'. Performing quantitative volumetric concrete or metal radio-analyses safer and faster (without lab intervention) was a key objective of this dynamic characterization approach. Currently, concrete core bores are shipped to certified laboratories where the concrete residue is run through a battery of tests to determine the contaminants. The existing core boring operation volatilises or washes out some of the contaminants (like tritium) and oftentimes cross-contaminates the are a around the core bore site. The volatilization of the contaminants can lead to airborne problems in the immediate vicinity of the core bore. Cross-contamination can increase the contamination area and thereby increase the amount of waste generated that needs to be treated and stabilized before disposal. The goal was to avoid those field activities that could cause this type of release. Therefore

  20. Radiation dose rates from commercial PWR and BWR spent fuel elements

    SciTech Connect

    Willingham, C.E.

    1981-10-01

    Data on measurements of gamma dose rates from commercial reactor spent fuel were collected, and documented calculated gamma dose rates were reviewed. As part of this study, the gamma dose rate from spent fuel was estimated, using computational techniques similar to previous investigations into this problem. Comparison of the measured and calculated dose rates provided a recommended dose rate in air versus distance curve for PWR spent fuel.

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

    SciTech Connect

    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.

  2. Crack growth rates of nickel alloy welds in a PWR environment.

    SciTech Connect

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.; Energy Technology

    2006-05-31

    In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320 C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of {approx}5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking.

  3. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    SciTech Connect

    J.S. Tang

    2001-05-03

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

  4. Three-dimensional printed knotted reactors enabling highly sensitive differentiation of silver nanoparticles and ions in aqueous environmental samples.

    PubMed

    Su, Cheng-Kuan; Hsieh, Meng-Hsuan; Sun, Yuh-Chang

    2016-03-31

    Whether silver nanoparticles (AgNPs) persist or release silver ions (Ag(+)) when discharged into a natural environment has remained an unresolved issue. In this study, we employed a low-cost stereolithographic three-dimensional printing (3DP) technology to fabricate the angle-defined knotted reactors (KRs) to construct a simple differentiation scheme for quantitative assessment of Ag(+) ions and AgNPs in municipal wastewater samples. We chose xanthan/phosphate-buffered saline as a dispersion medium for in situ stabilization of the two silver species, while also facilitating their extraction from complicated wastewater matrices. After method optimization, we measured extraction efficiencies of 54.5 and 32.3% for retaining Ag(+) ions and AgNPs, respectively, in the printed KR (768-turn), with detection limits (DLs) of 0.86 and 0.52 ng L(-1) when determining Ag(+) ions and AgNPs, respectively (sample run at pH 11 without a rinse solution), and 0.86 ng L(-1) when determining Ag(+) ions alone (sample run at pH 12 with a 1.5-mL rinse solution). The proposed scheme is tolerant of the wastewater matrix and provides more reliable differentiation between Ag(+)/AgNPs than does a conventional filtration method. The concept and applicability of adopting 3DP technology to renew traditional KR devices were evidently proven by means of these significantly improved analytical performance. Our analytical data suggested that the concentrations of Ag(+) ions and AgNPs in the tested industrial wastewater sample were both higher than those in domestic wastewater, implying that industrial activity might be a main source of environmental silver species, rather than domestic discharge from AgNP-containing products. PMID:26965333

  5. Adaptation of triple axis neutron spectrometer for SANS measurements using alumina samples at TRIGA reactor of Bangladesh

    NASA Astrophysics Data System (ADS)

    Ahmed, F. U.; Kamal, I.; Yunus, S. M.; Datta, T. K.; Azad, A. K.; Zakaria, A. K. M.; Goyal, P. S.

    2005-09-01

    Double crystal method known as Bonse and Hart's technique has been employed to develop small angle neutron scattering (SANS) facility on a triple axis neutron spectrometer at TRIGA Mark II (3 MW) research reactor of Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. Two Si(1 1 1) crystals with very small mosaic spread ∼1 min have been used for this purpose. At an incident neutron wavelength of 1.24 Å, this device is useful for SANS in the Q range between 1.6×10 -3 and 10 -1 Å -1. This Q range allows investigating particle sizes and interparticle correlations on a length scale of ∼200 Å. Results of SANS experiments on three alumina (Al 2O 3) samples as performed using above setup are presented. It is seen that Al 2O 3 particles, indeed, scatter neutrons in regions of small angles. It is also seen that scattering is different for different samples showing that it changes with a change in particle size.

  6. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    SciTech Connect

    Miro, R.; Maggini, F.; Barrachina, T.; Verdu, G.; Gomez, A.; Ortego, A.; Murillo, J. C.

    2006-07-01

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  7. Pump and valve fastener serviceability in PWR nuclear facilities

    SciTech Connect

    Moisidis, N.T.; Ratiu, M.D.

    1996-02-01

    The results of several studies conducted on corrosion of carbon and low-alloy steels in borated water have shown that impingement of borated steam on ferritic steels or contact with a moist paste of boric acid can lead to high corrosion rates due to high local concentrations of boric acid on the surface. The corrosion process of the flange fasteners of pumps and valves is considered a material compatibility and equipment maintenance problem. Therefore, the nuclear utilities of pressurized water reactor (PWR) power plants can prevent this damage by implementing appropriate fastener steel replacement and extended inspections to detect and correct the cause of leakage. A 3-phase corrosion protection program is presented for implementation based on system operability, outage-related accessibility, and cost of fastener replacement versus maintenance frequency increase. A selection criterion for fastener material is indicated based on service limitation: preloading and metal temperature.

  8. Ultrasonic Backscattering in Polycrystalline Materials of Pwr Components

    NASA Astrophysics Data System (ADS)

    Chassignole, B.; Dupond, O.; Fouquet, T.; Rupin, F.

    2011-06-01

    The ultrasonic examination of metallic components of Pressurized Water Reactors (PWR) is an important challenge for the nuclear industry. During the past decades, EDF R&D has undertaken numerous studies in order to improve the NDT process on these applications and to help to their qualification. The present paper deals with the problem of the structural noise which can potentially disturbs the ultrasonic inspection. In particular, this study proposes a modeling approach to simulate the ultrasonic scattering due to coarse grain structures of polycrystalline materials. The methodology is based on the mixing of a grain scale description of the material and a 2D finite element code (ATHENA) developed by EDF to simulate the ultrasonic propagation in isotropic and anisotropic elastic media. The modeling results are compared to experimental acquisitions on mock-ups containing artificial defects.

  9. Comparative analysis of isotopic composition of spent fuel from Takahama-3 PWR PIE database using TRIPOLI-PEPIN code

    SciTech Connect

    Lee, Y. K.

    2006-07-01

    Evaluation of isotopic composition of spent nuclear fuel is essential for reactor physics and fuel cycle back-end applications. A TRIPOLI-PEPIN coupled depletion code, TR4PEP, has been developed to meet these requirements. It combines the continuous-energy Monte Carlo transport code, TRIPOLI4.3 [1] and the point depletion code, PEPIN-2 [2], to perform the burnup dependent material data calculation. The depletion calculation flow of TR4PEP code has been presented on a previous study. Its application on PWR UO{sub 2} and MOX spent fuel has been validated against several international numerical benchmarks. Compared to industry standard deterministic cell codes and other Monte Carlo based depletion codes, TR4PEP deep-burn depletion calculations have shown satisfactory results. [3] In addition to the numerical benchmarks, the analysis of available post irradiation examination (PIE) results by TR4PEP is also important The PIE results at fuel assembly level are accessible only from spent fuel reprocessing plant and these data are not easy to use for code validation due to the dissolution of several assemblies in the same time. The PIE results at fuel pellet level depend not only on the method for the isotopic measurements but also on the irradiation environment and history. A free access PIE database on isotopic composition of spent nuclear fuel is obtainable from OECD/NEA. [4] Both PWR and BWR PIE data at fuel pellet level are taken into account in this database but the only 17 x 17 type PWR fuel available in this database is from Takahama-3 PIE results. To validate TR4PEP with Takahama-3 PIE results, two irradiated UO{sub 2} samples, SF95-4 from fuel assembly NT3G23 and SF97-5 from NT3G24, are considered in this study. Both samples have an initial {sup 235}U enrichment of 4.11 wt% and their burnup are respectively 36.69 and 47.03 GWd/t. Comparative analysis of isotopic composition from SF95-4 and SF97-5 including 19 actinides from {sup 234}U to {sup 247}Cm and 18

  10. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGESBeta

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    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

  11. Experience in PWR and BWR mixed-oxide fuel management

    SciTech Connect

    Schlosser, G.J.; Krebs, W.; Urban, P. )

    1993-04-01

    Germany has adopted the strategy of a closed fuel cycle using reprocessing and recycling. The central issue today is plutonium recycling by the use of U-Pu mixed oxide (MOX) in pressurized water reactors (PWRs) and boiling water reactors (BWRs). The design of MOX fuel assemblies and fuel management in MOX-containing cores are strongly influenced by the nuclear properties of the plutonium isotopes. Optimized MOX fuel assembly designs for PWRs currently use up to three types of MOX fuel rods having different plutonium contents with natural uranium or uranium tailings as carrier material but without burnable absorbers. The MOX fuel assembly designs for BWRs use four to six rod types with different plutonium contents and Gd[sub 2]O[sub 3]/UO[sub 2] burnable absorber rods. Both the PWR and the BWR designs attain good burnup equivalence and compatibility with uranium fuel assemblies. High flexibility exists in the loading schemes relative to the position and number of MOX fuel assemblies in the reloads and in the core as a whole. The Siemens experience with MOX fuel assemblies is based on the insertion of 318 MOX fuel assemblies in eight PWRs and 168 in BWRs and pressurized heavy water reactors so far. The primary operating results include information on the cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX fuel assemblies.

  12. Real-time PCR array chip with capillary-driven sample loading and reactor sealing for point-of-care applications.

    PubMed

    Ramalingam, Naveen; Liu, Hao-Bing; Dai, Chang-Chun; Jiang, Yu; Wang, Hui; Wang, Qinghui; M Hui, Kam; Gong, Hai-Qing

    2009-10-01

    A major challenge for the lab-on-a-chip (LOC) community is to develop point-of-care diagnostic chips that do not use instruments. Such instruments include pumping or liquid handling devices for distribution of patient's nucleic-acid test sample among an array of reactors and microvalves or mechanical parts to seal these reactors. In this paper, we report the development of a primer pair pre-loaded PCR array chip, in which the loading of the PCR mixture into an array of reactors and subsequent sealing of the reactors were realized by a novel capillary-based microfluidics with a manual two-step pipetting operations. The chip is capable of performing simultaneous (parallel) analyses of multiple gene targets and its performance was tested by amplifying twelve different gene targets against cDNA template from human hepatocellular carcinoma using SYBR Green I fluorescent dye. The versatility and reproducibility of the PCR-array chip are demonstrated by real-time PCR amplification of the BNI-1 fragment of SARS cDNA cloned in a plasmid vector. The reactor-to-reactor diffusion of the pre-loaded primer pairs in the chip is investigated to eliminate the possibility of primer cross-contamination. Key technical issues such as PCR mixture loss in gas-permeable PDMS chip layer and bubble generation due to different PDMS-glass bonding methods are investigated. PMID:19421862

  13. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    SciTech Connect

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  14. Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

    PubMed

    Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

    2002-12-01

    This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium). PMID:12509057

  15. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect

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

    2015-09-01

    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 that 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

  16. In-Situ Creep Testing Capability for the Advanced Test Reactor

    SciTech Connect

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2012-09-01

    An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

  17. Quantification and local distribution of hydrogen within Zircaloy-4 PWR nuclear fuel cladding tubes at the nuclear microprobe of the Pierre Süe Laboratory from μ-ERDA

    NASA Astrophysics Data System (ADS)

    Raepsaet, C.; Bossis, Ph.; Hamon, D.; Béchade, J. L.; Brachet, J. C.

    2008-05-01

    Hydrogen content and its distribution in in-core materials of nuclear plants are known to have a strong influence on their behaviour, especially on their mechanical properties but also on their corrosion resistance. This point has to be largely investigated in the case of the nuclear fuel cladding (Zr based alloys) of pressurized water reactors (PWR). Two situations have been considered here, with regards to the hydrogen content and its spatial distribution within the thickness of the tubes: Irradiated fuel cladding tubes after a nominal period under working conditions in a PWR core. Non-irradiated fuel cladding previously exposed to conditions representative of an hypothetical "loss of coolant accident" scenario (LOCA). As far as micrometric distributions of H were required, μ-ERDA has been performed at the nuclear microprobe of the Pierre Süe Laboratory. This facility is fitted with two beam lines. In the first one, used for non-active sample analysis, the μ-ERDA configuration has been improved to reduce the limits of detection and the reliability of the results. The second one offers the unique feature of being dedicated to radioactive samples. We will present the nuclear microprobe and emphasize on the μ-ERDA configuration of the two beam lines. We will illustrate the performance of the setup by describing the results obtained for Zircaloy-4 cladding both on non-irradiated and irradiated samples.

  18. Optimization of small long-life PWR based on thorium fuel

    SciTech Connect

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

    2015-09-30

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

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

  20. Optimization of small long-life PWR based on thorium fuel

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

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

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

    SciTech Connect

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

    2014-09-30

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

  2. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    DOE PAGESBeta

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7more » × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain

  3. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    SciTech Connect

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn

  4. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Edmondson, P. D.; Miller, M. K.; Powers, K. A.; Nanstad, R. K.

    2016-03-01

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m-2 (E > 1 MeV), and inlet temperatures of ∼289 °C (∼552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m-3, this copper level was below the solubility limit. A number density of 2 × 1022 m-3 of Ni-, Mn- Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m-3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m-3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain boundary in the low fluence

  5. The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

    NASA Astrophysics Data System (ADS)

    Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

    2007-05-01

    A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH( T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields.

  6. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS.

    SciTech Connect

    A. K. MAJI; B. MARSHALL; ET AL

    2000-10-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  7. Containment integrity of SEP plants under combined loads. [PWR; BWR

    SciTech Connect

    Lo, T.; Nelson, T.A.; Chen, P.Y.; Persinko, D.; Grimes, C.

    1984-06-01

    Because the containment structure is the last barrier against the release of radioactivity, an assessment was undertaken to identify the design weaknesses and estimate the margins of safety for the SEP containments under the postulated, combined loading conditions of a safe shutdown earthquake (SSE) and a design basis accident (DBA). The design basis accident is either a loss-of-coolant accident (LOCA) or a main steam line break (MSLB). The containment designs analyzed consisted of three inverted light-bulb shaped drywells used in boiling water reactor (BWR) systems, and three steel-lined concrete containments and a spherical steel shell used in pressurized water reactor (PWR) systems. These designs cover a majority of the containment types used in domestic operating plants. The results indicate that five of the seven designs are adequate even under current design standards. For the remaining two designs, the possible design weaknesses identified were buckling of the spherical steel shell and over-stress in both the radial and tangential directions in one of the concrete containments near its base.

  8. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  9. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  10. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  11. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

    SciTech Connect

    Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

    2007-08-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

  12. Possible use of the small Argus reactor for activation analysis of samples from gold-bearing deposits

    SciTech Connect

    Drynkin, V.I.; Kerzin, A.L.; Khvostionov, V.E.

    1987-09-01

    The authors study the feasibility of using the small Argus reactor with its power output of up to 50 kW as a neutron source in field neutron activation analysis work. A wide variety of gold ore types were irradiated and analyzed in order to evaluate the sensitivity and detection limits of the analyzer and the activation parameters of the thermal and epithermal neutrons generated by the reactor. These limits were found to practically fully satisfy the requirements for ore prospecting and processing work. A table lists the detection limits not only for gold but for 22 additional elements, including uranium, thorium, and several rare earths.

  13. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR & BWR accident sequences

    SciTech Connect

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-08-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences.

  14. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

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

    SciTech Connect

    J.W. Davis

    1996-08-29

    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.

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

    SciTech Connect

    Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C; Murphy, Brian D; Mueller, Don

    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 files 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.

  17. Chromosome aberrations induced in human lymphocytes by U-235 fission neutrons: I. Irradiation of human blood samples in the "dry cell" of the TRIGA Mark II nuclear reactor.

    PubMed

    Fajgelj, A; Lakoski, A; Horvat, D; Remec, I; Skrk, J; Stegnar, P

    1991-11-01

    A set-up for irradiation of biological samples in the TRIGA Mark II research reactor in Ljubljana is described. Threshold activation detectors were used for characterisation of the neutron flux, and the accompanying gamma dose was measured by TLDs. Human peripheral blood samples were irradiated "in vitro" and biological effects evaluated according to the unstable chromosomal aberrations induced. Biological effects of two types of cultivation of irradiated blood samples, the first immediately after irradiation and the second after 96 h storage, were studied. A significant difference in the incidence of chromosomal aberrations between these two types of samples was obtained, while our dose-response curve fitting coefficients alpha 1 = (7.71 +/- 0.09) x 10(-2) Gy-1 (immediate cultivation) and alpha 2 = (11.03 +/- 0.08) x 10(-2) Gy-1 (96 h delayed cultivation) are in both cases lower than could be found in the literature. PMID:1962281

  18. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    SciTech Connect

    McGraw, C.; Ilas, G.

    2012-07-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  19. Analysis of loss of off-site power with a PWR at shutdown

    SciTech Connect

    Chu, T.L.; Yoon, W.H.; Fitzpatrick, R.G.

    1987-01-01

    In many probabilistic risk assessments (PRAs), loss of offsite power (LOOP) when a nuclear power plant is operating was found to be a significant contributor to core damage. The purpose of this study is to provide an analysis of a LOOP event that occurs while a pressurized water reactor (PWR) is shut down. The importance of such an analysis was recognized as part of a study to evaluate the core damage frequency due to a loss of decay heat removal (DHR) capability during an outage. When a PWR is in a shutdown condition, there are relatively few technical specification requirements on the operability of safety systems. In fact, some safety systems are intentionally disabled, i.e., the safety injection system and nonoperating charging pumps. Another problem when the reactor is shut down is that the reactor coolant system (RCS) may be partially drained and the steam generators may be unavailable. To determine the time available for operator actions, given that a LOOP occurs during shutdown and the DHR capability is lost, a simple thermal model has been developed. Similar calculations have been performed for other phases of refueling and maintenance outages. A total core damage frequency due to LOOP while the plant is in shutdown has been calculated to be 5.9 x 10/sup -6//yr. This is approximately twice the core damage frequency calculated for LOOP when the plant is at power.

  20. Application of the MELCOR code to design basis PWR large dry containment analysis.

    SciTech Connect

    Phillips, Jesse; Notafrancesco, Allen; Tills, Jack Lee

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  1. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect

    Mardiansah, Deby; Takaki, Naoyuki

    2010-06-22

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  2. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  3. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    SciTech Connect

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-10-03

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation.

  4. MELCOR analyses of severe accident scenarios in Oconee, a B&W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S.; Foulds, R.

    1993-03-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock & Wilcox (B&W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  5. MELCOR analyses of severe accident scenarios in Oconee, a B W PWR plant

    SciTech Connect

    Madni, I.K.; Nimnual, S. ); Foulds, R. )

    1993-01-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock Wilcox (B W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides.

  6. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  7. Application of LBB to high energy piping systems in operating PWR

    SciTech Connect

    Swamy, S.A.; Bhowmick, D.C.

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  8. PWR secondary water chemistry guidelines: Revision 3. Final report

    SciTech Connect

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239).

  9. Strategies for selecting optimal sampling and work-up procedures for analysing alkylphenol polyethoxylates in effluents from non-activated sludge biofilm reactors.

    PubMed

    Stenholm, Ake; Holmström, Sara; Hjärthag, Sandra; Lind, Ola

    2012-01-01

    Trace-level analysis of alkylphenol polyethoxylates (APEOs) in wastewater containing sludge requires the prior removal of contaminants and preconcentration. In this study, the effects on optimal work-up procedures of the types of alkylphenols present, their degree of ethoxylation, the biofilm wastewater treatment and the sample matrix were investigated for these purposes. The sampling spot for APEO-containing specimens from an industrial wastewater treatment plant was optimized, including a box that surrounded the tubing outlet carrying the wastewater, to prevent sedimented sludge contaminating the collected samples. Following these changes, the sampling precision (in terms of dry matter content) at a point just under the tubing leading from the biofilm reactors was 0.7% RSD. The findings were applied to develop a work-up procedure for use prior to a high-performance liquid chromatography-fluorescence detection analysis method capable of quantifying nonylphenol polyethoxylates (NPEOs) and poorly investigated dinonylphenol polyethoxylates (DNPEOs) at low microg L(-1) concentrations in effluents from non-activated sludge biofilm reactors. The selected multi-step work-up procedure includes lyophilization and pressurized fluid extraction (PFE) followed by strong ion exchange solid phase extraction (SPE). The yields of the combined procedure, according to tests with NP10EO-spiked effluent from a wastewater treatment plant, were in the 62-78% range. PMID:22519096

  10. 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.

  11. Three-dimensional analysis of thermal and fluid mixing in cold leg and downcomer of PWR geometries

    SciTech Connect

    Lyczkowski, R.W.; Miao, C.C.; Domanus, H.M.; Hull, J.R.; Sha, W.T.; Schmitt, R.C.

    1983-12-01

    This report describes the three-dimensional transient and steady-state computations using the COMMIX-1A computer code for the analysis of six (6) 1/5-scale thermal and fluid mixing experiments conducted at Creare, Inc. under EPRI sponsorship. The tests chosen for analyses emphasized the effects of vent valve flow, cold leg and high pressure injection (HPI) coolant flow rates, and HPI location and geometry. The COMMIX-1A computations will provide fluid temperatures and velocities in the belt-line region of the downcomer for assessment of boundary conditions for thermal stress analysis in the vessel walls. A realistic prediction for thermal and fluid mixing significantly helps establish what overcooling transients can lead to in pressurized thermal shock (PTS) events. Sample three-dimensional steady-state computations are presented for three (3) generic full-scale pressurized water reactors (PWR's) typical of Westinghouse (W), Combustion Engineering (CE), and Babcock and Wilcox (B and W) configurations as part of the code assessment.

  12. Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

    SciTech Connect

    Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.

    2006-07-01

    High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

  13. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  15. Investigation of aerosols released at high temperature from nuclear reactor core models

    NASA Astrophysics Data System (ADS)

    Pintér Csordás, A.; Matus, L.; Czitrovszky, A.; Jani, P.; Maróti, L.; Hózer, Z.; Windberg, P.; Hummel, R.

    2000-12-01

    Two experiments were performed to simulate severe reactor accident with air ingress into the hot reactor core. The model bundles contained nine PWR type fuel rods. Their cladding was pre-oxidised by argon-oxygen (test 1) and steam (test 2). The released aerosol was measured continuously by laser particle counters. Morphology and elemental composition of the aerosol particles were studied on samples collected by impactors and quartz filters. The highest aerosol release was detected at the steepest rise of the bundle temperature. A second increase of the aerosol release appeared at the cooling down period. Because of the higher maximum temperature at test 2 about two orders of magnitude more uranium was released than in test 1. The highest emission was found for tin at test 1 and for zirconium and iron at test 2.

  16. Studies of the behavior of a reactor neutron beam at the sample position of a diffractometer using silicon monochromators

    NASA Astrophysics Data System (ADS)

    Ahmed, F. U.; Ahsan, M. H.; Khan, Aysha A.; Kamal, I.; Awal, M. A.; Ahmad, A. A. Z.

    1992-02-01

    A computer program TISTA has been developed for calculation of different aspects of designing a double axis neutron spectrometer at the TRIGA Mark II research reactor of the Atomic Energy Research Establishment, Dhaka, Bangladesh. The mathematical algorithms used in this program are based on the formalisms used by Fischer, Sabine and Bacon. Angle and energy resolutions and flux density as functions of neutron wave length, beam collimation, crystal asymmetry and deviation from zero-Bragg-angle position for different silicon crystal planes (111, 220, 311) have been calculated.

  17. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment

    NASA Astrophysics Data System (ADS)

    Zhang, Litao; Wang, Jianqiu

    2014-03-01

    Stress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen. The crack growth rate (CGR) increased with increasing DO content in the simulated PWR primary water. The fracture surface exhibited typical intergranular stress corrosion cracking (IGSCC) characteristics.

  18. Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5

    Energy Science and Technology Software Center (ESTSC)

    1999-06-02

    CONTEMPT4/MOD6 describes the response of multicompartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user-supplied descriptions of compartments,more » inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. CONTEMPT4/MOD6 also provides analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation to accommodate degraded core type accidents.« less

  19. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  20. Deposition of cobalt on surface-treated stainless steel under PWR conditions

    SciTech Connect

    Lister, D.H.; Anderson, P.G.; Barry, B.J.; Lavoie, R.G. . Chalk River Nuclear Labs.)

    1989-10-01

    As part of an on-going program aimed at reducing radiation exposures in light water reactors, the modification of surfaces to minimize their propensity to pick up radioactivity under reactor conditions has been studied. This report describes how stainless steel specimens, surface-treated with a variety of processes, picked up Co-60 from high-temperature water under PWR conditions in a high-pressure loop. The build-up of activity was monitored on-line with a movable gamma spectrometer. Off-line counting at the end of the experiment established the absolute activity levels, and selective examinations with SEM and metallography characterized the surface condition of the exceptional specimens. The effectiveness of the surface treatments was gauged by fitting simple parabolae to the activity build-up data and comparing the coefficients with those obtained from untreated control specimens. 10 refs., 23 figs., 4 tabs.

  1. Advances in reactor physics education: Visualization of reactor parameters

    SciTech Connect

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-07-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  2. A one-step preparation method of monolithic enzyme reactor for highly efficient sample preparation coupled to mass spectrometry-based proteomics studies.

    PubMed

    Jiang, Shan; Zhang, Zichuan; Li, Lingjun

    2015-09-18

    Mass spectrometry (MS) coupled to sample preparation and separation techniques has become a primary tool for proteomics studies. However, due to sample complexity, it is often challenging to achieve fast and efficient sample preparation prior to MS analysis. In recent decades, monolithic materials have been developed not only as chromatographic media, but also as efficient solid supports for immobilizing multiple types of affinity reagents. Herein, the N-acryloxysuccinimide-co-acrylamide-co-N,N'-methylenebisacrylamide (NAS-AAm-Bis) monolith was fabricated within silanized 200 μm i.d. fused-silica capillaries and was used as an immobilized enzyme reactor (IMER). The column was conjugated with trypsin/Lys-C and Lys-N enzymes to allow enzymatic digestions to occur while protein mixture was loaded onto the IMER column followed by MS-based proteomics analysis. Similar MS signal and protein sequence coverage were observed using protein standard bovine serum albumin (BSA) compared to in-solution digestion. Furthermore, mouse serum, yeast, and human cell lysate samples were also subjected to enzymatic digestion by both IMER (in seconds to minutes) and conventional in solution digestion (overnight) for comparison in large-scale proteomics studies. Comparable protein identification results obtained by the two methods highlighted the potential of employing NAS-based IMER column for fast and highly efficient sample preparation for MS analysis in proteomics studies. PMID:26300481

  3. Field application of a planted fixed bed reactor (PFR) for support media and rhizosphere investigation using undisturbed samples from full-scale constructed wetlands.

    PubMed

    Barreto, A B; Vasconcellos, G R; von Sperling, M; Kuschk, P; Kappelmeyer, U; Vasel, J L

    2015-01-01

    This study presents a novel method for investigations on undisturbed samples from full-scale horizontal subsurface-flow constructed wetlands (HSSFCW). The planted fixed bed reactor (PFR), developed at the Helmholtz Center for Environmental Research (UFZ), is a universal test unit for planted soil filters that reproduces the operational conditions of a constructed wetland (CW) system in laboratory scale. The present research proposes modifications on the PFR original configuration in order to allow its operation in field conditions. A mobile device to obtain undisturbed samples from real-scale HSSFCW was also developed. The experimental setting is presented with two possible operational configurations. The first allows the removal and replacement of undisturbed samples in the CW bed for laboratory investigations, guaranteeing sample integrity with a mobile device. The second allows the continuous operation of the PFR and undisturbed samples as a fraction of the support media, reproducing the same environmental conditions outside the real-scale system. Investigations on the hydrodynamics of the adapted PFR were carried out with saline tracer tests, validating the proposed adaptation. Six adapted PFR units were installed next to full-scale HSSFCW beds and fed with interstitial liquid pumped from two regions of planted and unplanted support media. Fourteen points were monitored along the system, covering carbon fractions, nitrogen and sulfate. The results indicate the method as a promising tool for investigations on CW support media, rhizosphere and open space for studies on CW modeling, respirometry, kinetic parameters, microbial communities, redox potential and plant influence on HSSFCW. PMID:26247753

  4. Nano-cavities observed in a 316SS PWR Flux Thimble Tube Irradiated to 33 and 70 dpa

    SciTech Connect

    Edwards, Danny J.; Garner, Francis A.; Bruemmer, Stephen M.; Efsing, Pal G.

    2009-02-28

    The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290ºC and 70 dpa at 315ºC were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during post-irradiation slow strain rate testing in PWR water conditions.

  5. Assessment of the radionuclide composition of "hot particles" sampled in the Chernobyl nuclear power plant fourth reactor unit.

    PubMed

    Bondarkov, Mikhail D; Zheltonozhsky, Viktor A; Zheltonozhskaya, Maryna V; Kulich, Nadezhda V; Maksimenko, Andrey M; Farfán, Eduardo B; Jannik, G Timothy; Marra, James C

    2011-10-01

    Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) Unit 4 Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified, and the fuel burn-up in these samples was determined. A systematic deviation in the burn-up values based on the cesium isotopes in comparison with other radionuclides was observed. The studies conducted were the first ever performed to demonstrate the presence of significant quantities of 242Cm and 243Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from 241Am (and going higher) in comparison with the theoretical calculations. PMID:21878762

  6. Structural integrity of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  7. Modeling the electrochemistry of the primary circuits of light water reactors

    SciTech Connect

    Bertuch, A.; Macdonald, D.D.; Pang, J.; Kriksunov, L.; Arioka, K.

    1994-12-31

    To model the corrosion behaviors of the heat transport circuits of light water reactors, a mixed potential model (NTM) has been developed and applied to both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Using the data generated by the GE/UKEA-Harwell radiolysis model, electrochemical potentials (ECPs) have been calculated for the heat transport circuits of eight BWRs operating under hydrogen water chemistry (HWC). By modeling the corrosion behaviors of these reactors, the effectiveness of HWC at limiting IGSCC and IASCC can be determined. For simulating PWR primary circuits, a chemical-radiolysis model (developed by the authors) was used to generate input parameters for the MPM. Corrosion potentials of Type 304 and 316 SSs in PWR primary environments were calculated using the NTM and were found to be in good agreement with the corrosion potentials measured in the laboratory for simulated PWR primary environments.

  8. 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

  9. Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Romano, A.J.

    1980-06-01

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  10. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  11. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-13

    ... COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors... and operate integral pressurized water reactors (iPWR). This guidance applies to environmental reviews... purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555...

  12. Zebra: An advanced PWR lattice code

    SciTech Connect

    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 precision and a high efficiency. (authors)

  13. Experimental development of power reactor advanced controllers

    SciTech Connect

    Edwards, R.M.; Weng, C.K.; Lindsay, R.W.

    1992-06-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  14. Experimental development of power reactor advanced controllers

    SciTech Connect

    Edwards, R.M. . Dept. of Nuclear Engineering); Weng, C.K. . Dept. of Mechanical Engineering); Lindsay, R.W. )

    1992-01-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

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

    SciTech Connect

    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 this 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)

  16. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    SciTech Connect

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  17. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    SciTech Connect

    Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa; Xu, Yiban; Cao, Liping

    2016-01-01

    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  18. 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.

  19. 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.

  20. Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

    NASA Astrophysics Data System (ADS)

    Ga¨rtner, M.; Fischer, G.

    1987-06-01

    To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

  1. Experiment data report for LOFT anticipated transient-without-scram Experiment L9-3. [PWR

    SciTech Connect

    Bayless, P.D.; Divine, J.M.

    1982-05-01

    Selected pertinent and uninterpreted data from the third anticipated transient with multiple failures experiment (Experiment L9-3) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large (approx. 1000 MW(e)), commercial PWR operations. Experiment L9-3 simulated a loss-of-feedwater anticipated transient without scram. The loss-of-feedwater accident led to an increase in the primary coolant system temperature and pressure. Both the experiment power-operated relief valve (PORV) and safety relief valve opened and were able to limit and control the pressure transient. The plant was then recovered with the control rods still withdrawn by injecting 7200-ppM borated water, manually cycling the PORV and feeding and bleeding the steam generator.

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

    SciTech Connect

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    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 of 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)

  3. Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

    SciTech Connect

    Jones, R.B.; Bolton, C.J.

    1997-02-01

    Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impurity precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.

  4. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    SciTech Connect

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface 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 both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  5. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  6. Expeditious method to determine uranium in the process control samples of chemical plant separating (233)U from thoria irradiated in power reactors.

    PubMed

    Kedari, C S; Kharwandikar, B K; Banerjee, K

    2016-11-01

    Analysis of U in the samples containing a significant proportion of (232)U and high concentration of Th is of great concern. Transmutation of Th in the nuclear power reactor produces a notable quantity of (232)U (half life 68.9 years) along with fissile isotope (233)U. The decay series of (232)U is initiated with (228)Th (half life 1.9 year) and it is followed by several short lived α emitting progenies, (224)Ra, (220)Rn, (216)Po, (212)Bi and (212)Po. Even at the smallest contamination of (228)Th in the sample, a very high pulse rate of α emission is obtained, which is to be counted for the radiometric determination of [U]. A commercially available anionic type of extractant Alamine®336 is used to obtain the selective extraction of U from other alpha active elements and fission products present in the sample. Experimental conditions of liquid-liquid extraction (LLE) are optimized for obtaining maximum decontamination and recovery of U in the organic phase. The effect of some interfering ionic impurities in the sample on the process of separation is investigated. Depending on the level of the concentration of U in the samples, spectrophotometry or radiometry methods are adopted for its determination after separation by LLE. Under optimized experimental conditions, i.e. 5.5M HCl in the aqueous phase and 0.27M Alamin®336 in the organic phase, the recovery of U is about 100%, the decontamination factor with respect to Th is >2000 and the extraction of fission products like (90)Sr, (144)Ce and (134,137)Cs is negligible. The detection limit for [U] using α radiometry is 10mg/L, even in presence of >100g/L of Th in the sample. Accuracy and precision for the determination of U is also assessed. Reproducibility of results is within 5%. This method shows very good agreement with the results obtained by mass spectrometry. PMID:27591623

  7. Small Reactor for Semiconductor Production by Neutron Transmutation Doping

    SciTech Connect

    Obara, Toru; Hong, Liem Peng

    2010-06-22

    New concept of small size nuclear reactor is proposed for Neutron Transmutation Doping (NTD). The reactor core consists of conventional PWR type fuel elements with light water moderator/coolant unlike conventional research reactors. Graphite reflector is employed for large neutron irradiation volume. Silicon ingots are put into the reflector region for irradiation. Neutronic analysis results show that this concept has possibility to product large amount of silicon ingots which have large diameter. An optimal reactor design and its performance are shown as a result of analysis in the paper.

  8. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  9. CRACK GROWTH RESPONSE OF ALLOY 152 AND 52 WELD METALS IN SIMULATED PWR PRIMARY WATER

    SciTech Connect

    Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2009-12-01

    The crack growth response of alloy 152 and 52 weld metals has been measured in simulated PWR primary water at both high (325-350 C) and low (50 C) temperatures. Tests were performed on samples machined from alloy 152 or 52 mockup welds. Propagation rates under cycle + hold and constant K conditions at high temperatures show stable, but extremely low SCC growth rates. The most significant intergranular cracking occurred during cycling at 50 C, particularly for the alloy 152 weld metal at high stress intensity.

  10. Fog inerting effects on hydrogen combustion in a PWR ice condenser contaminant

    SciTech Connect

    Luangdilok, W.; Bennett, R.B.

    1995-05-01

    A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H{sub 2}-air-steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser contaminant during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H{sub 2} by volume). 18 refs., 3 tabs.

  11. Calculation of the neutron source distribution in the VENUS PWR Mockup Experiment

    SciTech Connect

    Williams, M.L.; Morakinyo, P.; Kam, F.B.K.; Leenders, L.; Minsart, G.; Fabry, A.

    1984-01-01

    The VENUS PWR Mockup Experiment is an important component of the Nuclear Regulatory Commission's program goal of benchmarking reactor pressure vessel (RPV) fluence calculations in order to determine the accuracy to which RPV fluence can be computed. Of particular concern in this experiment is the accuracy of the source calculation near the core-baffle interface, which is the important region for contributing to RPV fluence. Results indicate that the calculated neutron source distribution within the VENUS core agrees with the experimental measured values with an average error of less than 3%, except at the baffle corner, where the error is about 6%. Better agreement with the measured fission distribution was obtained with a detailed space-dependent cross-section weighting procedure for thermal cross sections near the core-baffle interface region. The maximum error introduced into the predicted RPV fluence due to source errors should be on the order of 5%.

  12. COBRA/TRAC analysis of the PKL reflood test K9. [PWR

    SciTech Connect

    Wilkins, C.A.; Thurgood, M.J.

    1982-08-01

    Experiments at the Primaerkreislaeufe (PKL) test facility in Erlangen, Germany, simulated the refill and reflood procedure after a loss-of-coolant accident (LOCA) in the primary coolant system of a 1300-MW pressurized water reactor (PWR). COBRA/TRAC, a thermal-hydraulics analysis code developed at the Pacific Northwest Laboratory, was used to model experiment K9 of the PKL test series (completed December 1979). The COBRA/TRAC code, which utilizes COBRA-TF as the vessel module and TRAC-P1A for the remaining components, was designed to analyze LOCAs in PWRs. PKL-K9 was characterized by a double-ended guillotine break in the cold leg with emergency core cooling water injected into the cold legs. COBRA/TRAC was able to successfully predict lower-core temperature profiles and quench times, upper-core temperature profiles until the quench, upper plenum and break pressures, and correct trends in collapsed water levels.

  13. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    SciTech Connect

    Pasichnyk, I.; Perin, Y.; Velkov, K.

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  14. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    SciTech Connect

    Simion, G.P.; VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B.; Bulmahn, K.D.

    1993-06-01

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement.

  15. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    SciTech Connect

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  16. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  17. Effect of aging on the PWR Chemical and Volume Control System

    SciTech Connect

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K.

    1995-06-01

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  18. 3D Neutron Transport PWR Full-core Calculation with RMC code

    NASA Astrophysics Data System (ADS)

    Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien

    2014-06-01

    Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.

  19. 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.

  20. Automation of the quantitative determination of elemental content in samples using neutron activation analysis on the IBR-2 reactor at the frank laboratory for neutron physics, joint institute for nuclear research

    NASA Astrophysics Data System (ADS)

    Dmitriev, A. Yu.; Pavlov, S. S.

    2013-01-01

    Software for the automated quantitative determination of element concentrations in samples is described. This software is used in neutron activation analysis (NAA) at the IBR-2 reactor of the Frank Laboratory for Neutron Physics, Joint Institute for Nuclear Research (FLNP JINR).

  1. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  2. 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.

  3. Probabilistic analyses of failure in reactor coolant piping. [Double-ended guillotine break

    SciTech Connect

    Holman, G.S.

    1984-07-20

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB).

  4. 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

    SciTech Connect

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-09-01

    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, 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.

  5. Conceptual design of the advanced marine reactor MRX

    NASA Astrophysics Data System (ADS)

    1991-02-01

    Design studies on the advanced marine reactors have been done continuously since 1983 at the Japan Atomic Energy Research Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type pressurized water reactor (PWR) built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study.

  6. Actinide transmutation in a thermal reactor

    SciTech Connect

    Facchini, A.; Sanjust, V.

    1993-12-31

    The long term radiotoxicity of nuclear wastes may be substantially reduced by long irradiation in thermal reactors. Preliminary calculations showed that appreciable quantities of the minor actinides and long lived fission products may be recycled in a power PWR, and that, a few centuries after 20--30 years of irradiation, they reach radiotoxicity levels comparable to those of the uranium quantity required to make the corresponding fuel amount. The purpose of the present work is to investigate the conceptual possibility of reducing the level of the long term radiotoxicity, due to Minor Actinides and Long-Lived Fission Products (MA/LLFP) produced in UO{sub 2} fuel, by long irradiation of them in a power PWR. More precisely the authors pursued the objective of determining what fraction of the MA/LLFP mixture produced in a 1,000 MWe PWR during its whole life, may be burned in a similar power reactor. A waste burning efficiency has been considered satisfactory if the long term radiotoxicity of the MA/LLFP contained in a given quantity of spent fuel reaches, a few centuries after its irradiation, the level corresponding to that of the amount of natural uranium required to produce the same quantity of fresh fuel. This waiting time is in fact necessary in any case for cooling the other fission products to a sufficiently low radioactivity level and is a time span not unreasonable when considering man-made barriers against the radionuclide diffusion into the biosphere.

  7. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  8. Method of characteristics - Based sensitivity calculations for international PWR benchmark

    SciTech Connect

    Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

    2013-07-01

    Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

  9. Code System to Calculate Reactor Coolant System Leak Rate.

    Energy Science and Technology Software Center (ESTSC)

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report ofmore » the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.« less

  10. Code System to Calculate Reactor Coolant System Leak Rate.

    SciTech Connect

    Bell, Pat

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report of the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.