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Sample records for bwr mark-i containment

  1. Quantification of uncertainties in source term estimates for a BWR with Mark I containment

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Cazzoli, E.; Davis, R.; Ishigami, T.; Lee, M.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.

    1988-01-01

    A methodology for quantification and uncertainty analysis of source terms for severe accident in light water reactors (QUASAR) has been developed. The objectives of the QUASAR program are (1) to develop a framework for performing an uncertainty evaluation of the input parameters of the phenomenological models used in the Source Term Code Package (STCP), and (2) to quantify the uncertainties in certain phenomenological aspects of source terms (that are not modeled by STCP) using state-of-the-art methods. The QUASAR methodology consists of (1) screening sensitivity analysis, where the most sensitive input variables are selected for detailed uncertainty analysis, (2) uncertainty analysis, where probability density functions (PDFs) are established for the parameters identified by the screening stage and propagated through the codes to obtain PDFs for the outputs (i.e., release fractions to the environment), and (3) distribution sensitivity analysis, which is performed to determine the sensitivity of the output PDFs to the input PDFs. In this paper attention is limited to a single accident progression sequence, namely; a station blackout accident in a BWR with a Mark I containment buildings. Identified as an important accident in the draft NUREG-1150 a station blackout involves loss of both off-site power and DC power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation coding systems

  2. Direct torus venting analysis for Chinshan BWR-4 plant with MARK-I containment

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2017-03-15

    Highlights: • Study the effectiveness of Direct Torus Venting System (DTVS) during extended SBO of 24 h for Chinshan MARK-I plant. • Containment response is analyzed by GOTHIC based on boundary conditions from RETRAN calculation. • Analyses are performed with and without DTVS, respectively. • Suppression pool is sub-divided and thermal stratification is observed. - Abstract: The Chinshan plant, owned by Taiwan Power Company, has twin units of BWR-4 reactor and MARK-I containment. Both units have been operating at rated core thermal power of 1840 MWt. The existing Direct Torus Venting System (DTVS) is the main system used for venting the containment during the extended station blackout event. The purpose of this paper is to study the effects of the DTVS venting on the response of the containment pressure and temperature. The reactor is depressurized by manually opening the safety relief valves (SRVs) during the SBO, which causes the mass and energy to be discharged into and heat up the suppression pool. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The DTVS model is established in the GOTHIC model based on the venting size, venting piping loss, venting initiation time, and venting source. The lumped volume model, 1-D coarse-mesh model, and 3-D coarse-mesh model are considered in the torus volume. The calculation is first done without DTVS venting to establish a reference basis. Then a case with DTVS available is performed. Comparison of the two cases shows that the existing DTVS design is effective in mitigating the severity of the containment pressure and temperature transients. The results also show that the 1-D coarse-mesh model may not be appropriate since a

  3. Identification and assessment of containment and release management strategies for a BWR Mark I containment

    International Nuclear Information System (INIS)

    Lin, C.C.; Lehner, J.R.

    1991-09-01

    This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during the course of a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenges. A safety objective tree is developed which provides the connection between the safety objectives, the safety functions, the challenges, and the strategies. The strategies were assessed by applying them to certain severe accident sequence categories which have one or more of the following characteristics: have high probability of core damage or high consequences, lead to a number of challenges, and involve the failure of multiple systems. 59 refs., 55 figs., 27 tabs

  4. Containment venting as a mitigation technique for BWR MARK I plant ATWS

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1987-01-01

    Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure

  5. Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials

    International Nuclear Information System (INIS)

    Kress, T.S.; Cleveland, J.C.

    1988-01-01

    This study was undertaken to evaluate the thermal response of a BWR Mark-I containment shell in the event of an accident severe enough for molten core materials to fall into the cavity beneath the rector vessel and eventually come into direct contact with the shell. An existing ORNL three-dimensional transient heat transport computer code, HEATING-6, was used for a specific 2-D case (and variations) for which representative melt/shell boundary conditions required as input were available from other studies. In addition to the use of HEATING-6, a simplified analytical steady-state correlation was developed and given the name BWR Liner Analysis Program (BWRLAP). BWRLAP was ''benchmarked'' by comparison with HEATING-6 and was then used to make a number of parametric calculations to investigate the sensitivities of the results to the inputs. 5 refs., 11 figs., 2 tabs

  6. Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

    International Nuclear Information System (INIS)

    Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.; Khatib-Rahbar, M.

    1988-01-01

    In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probability Density Functions) associated with the radiological releases into the environment for the nine fission product groups at 10 hours following the initiation of core-concrete interactions. Also shown are the results ofthe STCP base case simulation. 5 refs., 1 fig., 1 tab

  7. Accident management information needs for a BWR with a MARK I containment

    Energy Technology Data Exchange (ETDEWEB)

    Chien, D.N.; Hanson, D.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1991-05-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs.

  8. Accident management information needs for a BWR with a MARK I containment

    International Nuclear Information System (INIS)

    Chien, D.N.; Hanson, D.J.

    1991-05-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs

  9. Mark I containment, short term program. Safety evaluation report

    International Nuclear Information System (INIS)

    1977-12-01

    Presented is a Safety Evaluation Report (SER) prepared by the Office of Nuclear Reactor Regulation addressing the Short Term Program (STP) reassessment of the containment systems of operating Boiler Water Reactor (BWR) facilities with the Mark I containment system design. The information presented in this SER establishes the basis for the NRC staff's conclusion that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety of the public, during an interim period of approximately two years while a methodical, comprehensive Long Term Program (LTP) is conducted. This SER also provides one of the basic foundations for the NRC staff review of the Mark I containment systems for facilities not yet licensed for operation

  10. Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

    International Nuclear Information System (INIS)

    Hyman, C.R.

    1988-01-01

    Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

  11. BWR Mark I pressure suppression study: bench mark experiments

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1977-01-01

    Computer simulations representative of the wetwell of Mark I BWR's have predicted pressures and related phenomena. However, calculational predictions for purposes of engineering decision will be possible only if the code can be verified, i.e., shown to compute in accord with measured values. Described in the report is a set of single downcomer spherical flask bench mark experiments designed to produce quantitative data to validate various air-water dynamic computations; the experiments were performed since relevant bench mark data were not available from outside sources. Secondary purposes of the study were to provide a test bed for the instrumentation and post-experiment data processing techniques to be used in the Laboratory's reactor safety research program and to provide additional masurements for the air-water scaling study

  12. Analyses of corium spreading in Mark I containment geometry

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Chu, C.C.; Farmer, M.T.

    1991-01-01

    An assessment of melt spreading in the Mark I system has been carried out using the MELTSPREAD-1 computer code together with supporting analyses. Application of MELTSPREAD-1 confirms the calculation of shell survival in a wet containment for the most probable melt release conditions from NUREG/CR-5423. According to MELTSPREAD-1, a dry containment also may not be threatened by melt spreading. This reflects the heat losses undergone by the melt in the process of spreading to the shell conservatively neglected in NUREG/CR-5423. However, there exist parameter ranges outside the most probable set where shell failure may be calculated. Accounting for the breakup and quenching of melt relocating through a deep layer of subcooled water also conservatively neglected in NUREG/CR-5423 can reduce the set of parameter variations for which containment failure is calculated in the wet case

  13. Severe accident progression perspectives for Mark I containments based on the IPE results

    International Nuclear Information System (INIS)

    Lin, C.C.; Lehner, J.R.; Pratt, W.T.; Drouin, M.

    1995-01-01

    Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe accident progresssion, insights obtained for (BWR) Mark I containments are discussed here. Mark I containments have high strength but small volumes and rely on pressure suppression pools to condense steam released from the reactor coolant system during an accident. Accidents causing structural failure of the drywell shortly after the core debris melts through the reactor vessel were found to be dominant contributors to risk. Importance of individual containment failure mechanisms depends on plant features and in some cases on modeling assumptions; however the following mechanisms were found important: drywell shell melt-through caused by direct contact with core debris and drywell failure caused by rapid pressure/temperature pulses at time of vessel melt-through. Drywell failure caused by gradual pressure/temperature buildup due to gases and steam released during core/concrete interactions is important in some IPEs. In other IPEs vent was an important contributor. However, accidents that bypass containment (eg interfacing systems LOCA)or involve containment isolation failure were not important contributors to the CDF in any of the IPEs for Mark I plants. These accidents are also not important to risk (even though they can involve large fission product release) because their frequencies of occurrence are so much lower than frequencies of early structural failure caused by other accidents that dominate the CDF

  14. The probability of Mark-I containment failure by melt-attack of the liner

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Yan, H.; Podowski, M.Z.

    1993-11-01

    This report is a followup to the work presented in NUREG/CR-5423 addressing early failure of a BWR Mark I containment by melt attack of the liner, and it constitutes a part of the implementation of the Risk-Oriented Accident Analysis Methodology (ROAAM) employed therein. In particular, it expands the quantification to include four independent evaluations carried out at Rensselaer Polytechnic Institute, Argonne National Laboratories, Sandia National Laboratories and ANATECH, Inc. on the various portions of the phenomenology involved. These independent evaluations are included here as Parts II through V. The results, and their integration in Part I, demonstrate the substantial synergism and convergence necessary to recognize that the issue has been resolved

  15. Mark I containment long-term program safety evaluation report, resolution of generic technical activity A-7. Report for February 1977-December 1979

    International Nuclear Information System (INIS)

    1980-07-01

    During testing for an advanced Boiling Water Reactor (BWR) containment system design (Mark III), suppression pool hydrodynamic loads were identified which had not been considered in the original design of the Mark I containment system. To address this issue, a Mark I Owners Group was formed and the assessment was divided into a short-term and long-term program. The results of the NRC staff's review of the Mark I Containment Short Term Program are described in NUREG-0408. This report describes the results of the NRC staff's review of the generic Mark I Containment Long Term Program (LTP). The LTP was conducted to provide a generic basis to define suppression pool hydrodynamic loads and the related structural acceptance criteria, such that a comprehensive reassessment of each Mark I containment system would be performed. A series of experimental and analytical programs were conducted by the Mark I Owners Group to provide the necessary bases for the generic load definition and structural assessment techniques. The generic methods proposed by the Mark I Owners Group, as modified by the NRC staff's requirements, will be used to perform plant-unique analyses, which will identify the plant modifications, if any, that will be needed to restore the originally intended margin of safety in the Mark I containment designs

  16. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yen-Shu, E-mail: yschen@iner.org.t [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

    2011-05-15

    Research highlights: The Chinshan Mark I containment pressure-temperature responses are analyzed. GOTHIC is used to calculate the containment responses under three pipe break events. This study is used to support the Chinshan Stretch Power Uprate (SPU) program. The calculated peak pressure and temperature are still below the design values. The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 {sup o}C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 {sup o}C). Additionally, the peak drywell temperature of 155.3 {sup o}C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 {sup o}C, which is below the pool temperature used for evaluating the

  17. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon

    2011-01-01

    Research highlights: → The Chinshan Mark I containment pressure-temperature responses are analyzed. → GOTHIC is used to calculate the containment responses under three pipe break events. → This study is used to support the Chinshan Stretch Power Uprate (SPU) program. → The calculated peak pressure and temperature are still below the design values. → The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 o C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 o C). Additionally, the peak drywell temperature of 155.3 o C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 o C, which is below the pool temperature used for evaluating the

  18. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

  19. Human reliability analysis for venting a BWR Mark I during a severe accident

    International Nuclear Information System (INIS)

    Nelson, W.R.; Blackman, H.S.

    1986-01-01

    A Human Reliability Analysis (HRA) was performed for the operator actions necessary to achieve containment venting for the Peach Bottom Atomic Power Station. This study was funded by the United States Nuclear Regulatory Commission (USNRC) and performed by the Idaho National Engineering Laboratory (INEL). The goal of the analysis was to estimate Human Error Probabilities (HEPs) to determine the likelihood that operators would fail to complete the venting process. The analysis was performed for two generic accident sequences: anticipated transient without scram (ATWS) and station blackout. Two major methods were used to estimate the HEPs: Technique for Human Error rate Prediction (THERP) and Success Likelihood Index Methodology (SLIM). For the ATWS scenarios analyzed, the calculated HEPs ranged from 0.23 to 0.35, depending on the number of vent paths that are required to reduce the containment pressure. It should be noted that the confidence bounds around these HEPs are large, However, even when considering the large confidence range, the failure probabilities are larger than what is typical for normal operator actions. For station blackout, the HEP is 1.0, resulting from the dangerous environmental conditions that are present, assuming that plant management would not deliberately expose personnel to a potentially fatal environment. These results are based on the analysis of draft procedures for containment venting. It is probable that careful revision of the procedures could reduce the human error probabilities

  20. External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure - Scoping Assessment

    International Nuclear Information System (INIS)

    Robb, Kevin R.

    2017-01-01

    This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

  1. External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

  2. BWR MARK I pressure suppression pool mixing and stratification analysis using GOTHIC lumped parameter modeling methodology

    International Nuclear Information System (INIS)

    Ozdemir, Ozkan Emre; George, Thomas L.

    2015-01-01

    As a part of the GOTHIC (GOTHIC incorporates technology developed for the electric power industry under the sponsorship of EPRI.) Fukushima Technical Evaluation project (EPRI, 2014a, b, 2015), GOTHIC (EPRI, 2014c) has been benchmarked against test data for pool stratification (EPRI, 2014a, b, Ozdemir and George, 2013). These tests confirmed GOTHIC’s ability to simulate pool mixing and stratification under a variety of anticipated suppression pool operating conditions. The multidimensional modeling requires long simulation times for events that may occur over a period of hours or days. For these scenarios a lumped model of the pressure suppression chamber is desirable to maintain reasonable simulation times. However, a lumped model for the pool is not able to predict the effects of pool stratification that can influence the overall containment response. The main objective of this work is on the development of a correlation that can be used to estimate pool mixing and stratification effects in a lumped modeling approach. A simplified lumped GOTHIC model that includes a two zone model for the suppression pool with controlled circulation between the upper and lower zones was constructed. A pump and associated flow connections are included to provide mixing between the upper and lower pool volumes. Using numerically generated data from a multidimensional GOTHIC model for the suppression pool, a correlation was developed for the mixing rate between the upper and lower pool volumes in a two-zone, lumped model. The mixing rate depends on the pool subcooling, the steam injection rate and the injection depth

  3. BWR steel containment corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Tan, C.P.; Bagchi, G.

    1996-04-01

    The report describes regulatory actions taken after corrosion was discovered in the drywell at the Oyster Creek Plant and in the torus at the Nine Mile Point 1 Plant. The report describes the causes of corrosion, requirements for monitoring corrosion, and measures to mitigate the corrosive environment for the two plants. The report describes the issuances of generic letters and information notices either to collect information to determine whether the problem is generic or to alert the licensees of similar plants about the existence of such a problem. Implementation of measures to enhance the containment performance under severe accident conditions is discussed. A study by Brookhaven National Laboratory (BNL) of the performance of a degraded containment under severe accident conditions is summarized. The details of the BNL study are in the appendix to the report.

  4. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  5. BWR Mark I pressure suppression study: characterization of the vertical load function utilizing bench top model tests

    International Nuclear Information System (INIS)

    McCauley, E.W.; Lai, W.

    1977-02-01

    A study was conducted to characterize the mechanisms which give rise to observed oscillations in the vertical load function (VLF) of bench top pool dynamics tests. This is part of a continuing investigation at the Lawrence Livermore Laboratory of the General Electric Mark I Nuclear Reactor pressure suppression system

  6. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  7. BWR containments license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Smith, S.; Gregor, F.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

  8. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

  9. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-01-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  10. Evaluation of JRC source term methodology using MAAP5 as a fast-running crisis tool for a BWR4 Mark I reactor

    International Nuclear Information System (INIS)

    Vela-García, M.; Simola, K.

    2016-01-01

    JRC participated in the OECD/NEA FASTRUN benchmark reviewing fast-running software tools to model fission product releases during accidents at nuclear power plants. The main goal of fast-running software tools is to foresee the accident progression, so that mitigating actions can be taken and the population can be adequately protected. Within the FASTRUN, JRC used the MAAP 4.0.8 code and developed a methodology to obtain the source term (as activity released per radioisotope) of PWR and BWR station black-out accident scenarios. The modifications made in the MAAP models were limited to a minimum number of important parameters. This aims at reproducing a crisis situation with a limited time to adapt a generic input deck. This paper presents further studies, where JRC analysed the FASTRUN BWR scenario using MAAP 5.0.2 that has the capability of calculating doses. A sensitivity study was performed with the MAAP 5.0.2 DOSE package deactivated, using the same methodology as in the case of MAAP 4.0.8 for source term calculation. The results were close to the reference LTSBO SOARCA case, independently of the methodology used. One of the benefits of using the MAAP code is the short runtime of the simulations.

  11. Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

    International Nuclear Information System (INIS)

    Su, T.M.

    1982-10-01

    Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report

  12. Computer-aided digitization of graphical mass flow data from the 1/5-scale Mark I BWR pressure suppression experiment

    International Nuclear Information System (INIS)

    Holman, G.S.; McCauley, E.W.

    1979-01-01

    Periodically in the analysis of engineering data, it becomes necessary to use graphical output as the solitary source of accurate numerical data for use in subsequent calculations. Such was our experience in the extended analysis of data from the 1/5-scale Mark I boiling water reactor pressure suppression experiment (PSE). The original numerical results of extensive computer calculations performed at the time of the actual PSE tests and required for the later extended analysis program had not been retained as archival records. We were, therefore, required to recover the previously calculated data, either by a complete recalculation or from available computer graphics records. Time constraints suggested recovery from the graphics records as the more viable approach. This report describes two different approaches to recovery of digital data from graphics records. One, combining hard and software techniques immediately available to us at LLL, proved to be inadequate for our purposes. The other approach required the development of pure software techniques that interfaced with LLL computer graphics to unpack digital coordinate information directly from graphics files. As a result of this effort, we were able to recover the required data with no significant loss in the accuracy of the original calculations

  13. The impact of BWR MK I primary containment failure dynamics on secondary containment integrity

    International Nuclear Information System (INIS)

    Greene, S.R.

    1987-01-01

    During the past four years, the ORNL BWRSAT Program has developed a series of increasingly sophisticated BWR secondary containment models. These models have been applied in a variety of studies to evaluate the severe accident mitigation capability of BWR secondary containments. This paper describes the results of a recent ORNL study of the impact of BWR MK I primary containment failure dynamics on secondary containment integrity. A 26-cell MELCOR Browns Ferry secondary containment model is described and the predicted thermodynamic response of the secondary containment to a variety of postulated primary containment failure modes is presented. The effects of primary containment failure location, timing, and ultimate hole size on secondary containment response is investigated, and the potential impact of hydrogen deflagrations on secondary containment integrity is explored

  14. Corrosion failure of a BWR embedded reactor containment liner

    International Nuclear Information System (INIS)

    Wegemar, B.

    2006-01-01

    Following sixteen fuel cycles, leakage through a BWR embedded reactor containment liner (carbon steel) was discovered. Leakage was located at a penetration for electrical conductors as a result of penetrating corrosion attack. During construction, porous cement structures and air pockets/cavities were formed due to erroneous injection of grout. Corrosion attacks on the CS steel liner were located at the relatively small, active surfaces in contact with the porous cement structure. The corrosion mechanism was supposed to be anodic dissolution of the steel liner in areas with insufficient passivation. The penetrations were restored according to original design requirements. (author)

  15. Design loads, loading combinations and structural acceptance criteria for BWR containments in the United States

    International Nuclear Information System (INIS)

    Edwards, N.W.

    1979-01-01

    The definition of loads, loading combinations, and structural acceptance criteria used for the design and evaluation of BWR containments in the Unites States has become much more comprehensive over the past decade. The Mark I pressure suppression containment vessels were designed for a static design pressure, a design temperature, dead load and static equivalent earthquake. The current Mark III containments are being designed to accommodate many more loads such as safety relief valve discharge loads, and suppression pool hydrodynamic loadings associated with the steam condensation phenomena as well as pressure and temperature transients for a range of pipe break sizes. Consistent with the more comprehensive definition of loads and loading combinations, the ASME Code presently establishes structural acceptance criteria with different margins of safety by the definition of Service Level Assignments A, B, C and D. Acting in a responsible manner, United States utilities are currently evaluating and modifying existing containment vessels to account for the more detailed load definition and structural acceptance criteria. (orig.)

  16. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  17. BWR Mark III pressure suppression containment response to hydrogen deflagration

    International Nuclear Information System (INIS)

    Fuls, G.M.; Gunter, A.D.

    1982-01-01

    The CLASIX-3 computer program has been used to evaluate the temperature and pressure response of the BWR Mark III Suppression Containment System to hydrogen deflagration resulting from a degraded core condition. The CLASIX-3 computer program is an extension of the CLASIX program which was originally developed to analyze ice condenser containments. A brief description is given of the modifications made to CLASIX to increase its flexibility and versatility to include the capability of analyzing the Mark III Containment. Analytical results are presented for the two base case transients. The two base cases are the stuck open steam relief valve and the small break LOCA, both of which are assumed to lead to a degraded core condition and the release of hydrogen to the containment. Results include pressure and temperature response, gas concentrations and suppression pool response

  18. Effect of torus wall flexibility on hydro-structural interaction in BWR containment system

    International Nuclear Information System (INIS)

    Lu, S.C.H.; McCauley, E.W.; Holman, G.S.

    1979-01-01

    The MARK I boiling water reactor (BWR) containment system is comprised of a light-bulb-shaped reactor compartment connected through vent pipes to a torus-shaped and partially water-filled pressure suppression chamber, or the wetwell. During either a normally occurring safety relief valve (SRV) discharge or a hypothetical loss-of-coolant accident (LOCA), air or steam is forced into the wetwell water pool for condensation and results in hydrodynamically induced loads on the torus shell. An analytical program is described which employs the finite element method to investigate the influence of torus wall flexibility on hydrodynamically induced pressure and the resultant force on the torus shell surface. The shell flexibility is characterized by the diameter-to-thickness ratio which is varied from the perfectly rigid case to the nominal plant condition. The general conclusion reached is that torus wall flexibility decreases both the maximum pressure seen by the shell wall and the total vertical load resulted from the hydrodynamically induced pressure

  19. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  20. Thermal hydraulic analysis of BWR containment venting system

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

    2015-01-01

    Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

  1. GOTHIC MODEL OF BWR SECONDARY CONTAINMENT DRAWDOWN ANALYSES

    International Nuclear Information System (INIS)

    Hansen, P.N.

    2004-01-01

    This article introduces a GOTHIC version 7.1 model of the Secondary Containment Reactor Building Post LOCA drawdown analysis for a BWR. GOTHIC is an EPRI sponsored thermal hydraulic code. This analysis is required by the Utility to demonstrate an ability to restore and maintain the Secondary Containment Reactor Building negative pressure condition. The technical and regulatory issues associated with this modeling are presented. The analysis includes the affect of wind, elevation and thermal impacts on pressure conditions. The model includes a multiple volume representation which includes the spent fuel pool. In addition, heat sources and sinks are modeled as one dimensional heat conductors. The leakage into the building is modeled to include both laminar as well as turbulent behavior as established by actual plant test data. The GOTHIC code provides components to model heat exchangers used to provide fuel pool cooling as well as area cooling via air coolers. The results of the evaluation are used to demonstrate the time that the Reactor Building is at a pressure that exceeds external conditions. This time period is established with the GOTHIC model based on the worst case pressure conditions on the building. For this time period the Utility must assume the primary containment leakage goes directly to the environment. Once the building pressure is restored below outside conditions the release to the environment can be credited as a filtered release

  2. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  3. Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Dixon, D.; Lord, D.

    1978-01-01

    The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

  4. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    Greene, S.R.

    1986-01-01

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

  5. Identification and assessment of containment and release management strategies for a BWR Mark III containment

    International Nuclear Information System (INIS)

    Lin, C.C.; Lehner, J.R.; Vandenkieboom, J.J.

    1992-02-01

    This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark III type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark III containment could face during the course of a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenges. The strategies are linked to the general safety objectives which apply for containment and release management by means of a safety objective tree. The strategies were assessed by applying them to certain severe accident sequence categories deemed important for a Mark III containment because of one or more of the following characteristics: high probability of core damage, high consequences, lead to a number of challenges, and involve the failure of multiple systems

  6. Role of BWR secondary containments in severe accident mitigation: issues and insights from recent analyses

    International Nuclear Information System (INIS)

    Greene, S.R.

    1988-01-01

    All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield building, auxiliary building and fuel building (MK III), or an auxiliary building and enclosure building (Grand Gulf style MK III). Although secondary containment designs are highly plant specific, their purpose is to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident. This paper presents a brief overview of domestic BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent ORNL secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented

  7. Analysis CFD for the hydrogen transport in the primary containment of a BWR

    International Nuclear Information System (INIS)

    Jimenez P, D. A.; Del Valle G, E.; Gomez T, A. M.

    2014-10-01

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  8. Analysis of radiological consequences in a typical BWR with a mark-II containment

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro

    2003-01-01

    INS/NUPEC in Japan has been carrying out the Level 3 PSA program. In the program, the MACCS2 code has been extensively applied to analyze radiological consequences for typical BWR and PWR plants in Japan. The present study deals with analysis of effects of the AMs, which were implemented by industries, on radiological consequence for a typical BWR with a Mark-II containment. In the present study, source terms and their frequencies of source terms were used based on results of Level 2 PSA taking into account AM countermeasures. Radiological consequences were presented with dose risks (Sv/ry), which were multiplied doses (Sv) by containment damage frequencies (/ry), and timing of radionuclides release to the environment. The results of the present study indicated that the dose risks became negligible in most cases taking AM countermeasures and evacuations. (author)

  9. An analysis of containment venting as a severe accident mitigation strategy for the BWR Mark II containment

    International Nuclear Information System (INIS)

    Kelly, D.L.; Galyean, W.J.

    1990-01-01

    An evaluation of a BWR/4 reactor with a Mark-II containment has identified the effects of containment venting on core damage frequency and containment failure mode, and has performed a limited evaluation of the effects on the off-site consequences. The analysis was founded upon an existing probabilistic risk assessment (PRA) with the addition of a proposed filtered containment venting system, based on the Swedish Filtra system installed at the Barseback nuclear power station in southern Sweden. Three different containment venting strategies were examined for their effects on plant risk. These are discussed

  10. Maintaining leak tightness capability of Caorso BWR containment

    International Nuclear Information System (INIS)

    Barsanti, P.; Di Palo, L.; Grimaldi, G.

    1988-01-01

    In 1987 the local leak rate test (LLRT) results of the primary containment were revised, with the following main goals: to highlight recurring problems, leading to lack of leak tightness of the primary containment; to individuate the pertinent degradation mechanisms; to assess the corrective actions already implemented and to plan further improvements, if necessary; and to optimize the preventive maintenance program on the containment, particularly the inspection frequency. All LLRTs in the past operating period, both before (as found) and after (as left) maintenance were analyzed, in terms of leakage rate and equivalent area of leak, for each penetration. Corrective actions already implemented included replacement of some valves with better quality type one, passivation of the carbon steel pipes and improvement of the pertinent surveillance procedures. Long term corrective actions, now under consideration, will include the following: more extensive passivation of pipes, carrying humid air, so that oxidation could be drastically reduced; better chemistry control in fluid systems; extensive replacement of the butterfly valves presently used; implementation of the LLRT practice, such to quantitatively measure the leakage rate, also in presence of large leak; and reduction of the time interval between periodical tests, on the basis of the results of the previous ones. Following these guidelines, future overall leakage tests would be performed in as found condition, aimed to verify the effectiveness of the entire maintenance and testing program of the primary containment and of its capability to maintain leak tightness during the time between two subsequent tests

  11. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  12. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor

    International Nuclear Information System (INIS)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L.; Tijerina S, F.; Tapia M, R.

    2016-09-01

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  13. Analysis CFD for the hydrogen transport in the primary containment of a BWR; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Gomez T, A. M., E-mail: guerreroazteca_69@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    This study presents a qualitative and quantitative comparison among the CFD GASFLOW and OpenFOAM codes which are related with the phenomenon of hydrogen transport and other gases in the primary containment of a Boiling Water Reactor (BWR). GASFLOW is a commercial license code that is well validated and that was developed in Germany for the analysis of the gases transport in containments of nuclear reactors. On the other hand, OpenFOAM is an open source code that offers several evaluation solvers for different types of phenomena; in this case, the solver reacting-Foam is used to analyze the hydrogen transport inside the primary containment of the BWR. The results that offer the solver reacting-Foam of OpenFOAM are evaluated in the hydrogen transport calculation and the results are compared with those of the program of commercial license GASFLOW to see if is viable the use of the open source code in the case of the hydrogen transport in the primary containment of a BWR. Of the obtained results so much quantitative as qualitative some differences were identified between both codes, the differences (with a percentage of maximum error of 4%) in the quantitative results are small and they are considered acceptable for this analysis type, also, these differences are attributed mainly to the used transport models, considering that OpenFOAM uses a homogeneous model and GASFLOW uses a heterogeneous model. (Author)

  14. Development of a parametric containment event tree model for a severe BWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T [OTO-Consulting Ay, Helsinki (Finland)

    1995-04-01

    A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

  15. BWR Mark III containment analyses using a GOTHIC 8.0 3D model

    International Nuclear Information System (INIS)

    Jimenez, Gonzalo; Serrano, César; Lopez-Alonso, Emma; Molina, M del Carmen; Calvo, Daniel; García, Javier; Queral, César; Zuriaga, J. Vicente; González, Montserrat

    2015-01-01

    Highlights: • The development of a 3D GOTHIC code model of BWR Mark-III containment is described. • Suppression pool modelling based on the POOLEX STB-20 and STB-16 experimental tests. • LOCA and SBO transient simulated to verify the behaviour of the 3D GOTHIC model. • Comparison between the 3D GOTHIC model and MAAP4.07 model is conducted. • Accurate reproduction of pre severe accident conditions with the 3D GOTHIC model. - Abstract: The purpose of this study is to establish a detailed three-dimensional model of Cofrentes NPP BWR/6 Mark III containment building using the containment code GOTHIC 8.0. This paper presents the model construction, the phenomenology tests conducted and the selected transient for the model evaluation. In order to study the proper settings for the model in the suppression pool, two experiments conducted with the experimental installation POOLEX have been simulated, allowing to obtain a proper behaviour of the model under different suppression pool phenomenology. In the transient analyses, a Loss of Coolant Accident (LOCA) and a Station Blackout (SBO) transient have been performed. The main results of the simulations of those transients were qualitative compared with the results obtained from simulations with MAAP 4.07 Cofrentes NPP model, used by the plant for simulating severe accidents. From this comparison, a verification of the model in terms of pressurization, asymmetric discharges and high pressure release were obtained. The completeness of this model has proved to adequately simulate the thermal hydraulic phenomena which occur in the containment during accidental sequences

  16. An overview of the BWR ECCS strainer blockage issues

    International Nuclear Information System (INIS)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-01-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

  17. Twenty years of Triga Mark I reactor use

    International Nuclear Information System (INIS)

    Stasiulevicius, R.; Maretti Junior, F.

    1981-01-01

    This work is a report on the 20 years of activities of the Triga Mark I, research reactor located in Belo Horizonte, Brazil. It contains also a list of publications, details of operation and improvements introduced in the reactor as well as some perspectives for its future. (A.C.A.S.)

  18. Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

    Energy Technology Data Exchange (ETDEWEB)

    Mueftueoglu, A.K.; Feltus, M.A. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and other signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.

  19. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng [Atomic Energy Council, Taoyuan City, Taiwan (China). Inst. of Nuclear Energy Research

    2017-11-15

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  20. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    International Nuclear Information System (INIS)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng

    2017-01-01

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  1. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  2. THE MARK I BUSINESS SYSTEM SIMULATION MODEL

    Science.gov (United States)

    of a large-scale business simulation model as a vehicle for doing research in management controls. The major results of the program were the...development of the Mark I business simulation model and the Simulation Package (SIMPAC). SIMPAC is a method and set of programs facilitating the construction...of large simulation models. The object of this document is to describe the Mark I Corporation model, state why parts of the business were modeled as they were, and indicate the research applications of the model. (Author)

  3. Two types of a passive safety containment for a near future BWR with active and passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

    2009-01-01

    The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

  4. Two types of a passive safety containment for a near future BWR with active and passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Takashi [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)], E-mail: takashi44.sato@glb.toshiba.co.jp; Akinaga, Makoto; Kojima, Yoshihiro [Toshiba Corporation, IEC, Gen-SS, 8, Shinsugita-ho, Isogo-ku, Yokohama (Japan)

    2009-09-15

    The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

  5. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  6. Identification and assessment of containment and release management strategies for a BWR Mark II containment

    International Nuclear Information System (INIS)

    Lin, C.C.; Lehner, J.R.

    1992-06-01

    Accident management strategies that have the potential to maintain containment integrity and control or mitigate the release of radioactivity following a severe accident at a boiling water reactor with a Mark 2 type of containment are identified and evaluated. The strategies are referred to as containment and release strategies. Using information available from probabilistic risk assessments and other existing severe accident research, and employing simplified containment and release event trees, this report identified the challenges a Mark 2 containment may encounter during a severe accident, the mechanisms behind these challenges, and the strategies that could be used to mitigate the challenge. By means of a safety objective tree, the strategies are linked to the general safety objectives of containment and release management. As part of the assessment process, the strategies are applied to certain severe accident sequence categories deemed important to a Mark 2 containment. These sequence categories exhibit one or more of the following characteristics: high probability of core damage, high consequences, lead to a number of challenges, and involve the failure of multiple systems. The Limerick Generating Station is used as a representative Mark 2 plant to illustrate plant specifics in this report

  7. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    Eckardt, B.

    1991-01-01

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO 2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  8. Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

    International Nuclear Information System (INIS)

    Wiles, L.E.; McCann, R.A.

    1981-09-01

    This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

  9. An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.; Podowski, M.Z.

    1991-01-01

    This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs

  10. An assessment of BWR [boiling water reactor] Mark-II containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J.; Wagner, K.C.

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs

  11. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

  12. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 1

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1981-01-01

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherlands and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.) [de

  13. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 2

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1981-01-01

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherland and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.) [de

  14. Containment performance improvement program

    International Nuclear Information System (INIS)

    Beckner, W.; Mitchell, J.; Soffer, L.; Chow, E.; Lane, J.; Ridgely, J.

    1990-01-01

    The Containment Performance Improvement (CPI) program has been one of the main elements in the US Nuclear Regulatory Commission's (NRC's) integrated approach to closure of severe accident issues for US nuclear power plants. During the course of the program, results from various probabilistic risk assessment (PRA) studies and from severe accident research programs for the five US containment types have been examined to identify significant containment challenges and to evaluate potential improvements. The five containment types considered are: the boiling water reactor (BMR) Mark I containment, the BWR Mark II containment, the BWR Mark III containment, the pressurized water reactor (PWR) ice condenser containment, and the PWR dry containments (including both subatmospheric and large subtypes). The focus of the CPI program has been containment performance and accident mitigation, however, insights are also being obtained in the areas of accident prevention and accident management

  15. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  16. Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Collins, E.K.; Lai, W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

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

    International Nuclear Information System (INIS)

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

    1996-01-01

    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

  18. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  19. CFD Simulation of rigid venting of the containment of a BWR-5 Mark-II reactor; Simulacion CFD de los venteos rigidos de la contencion de un reactor BWR-5 Mark-II

    Energy Technology Data Exchange (ETDEWEB)

    Galindo G, I. F.; Vazquez B, A. K.; Velazquez E, L. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F.; Tapia M, R., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

    2016-09-15

    In conditions of prolonged loss of external energy or a severe accident, venting to the atmosphere is an alternative to prevent overpressure and release of fission products from the primary containment of a nuclear reactor. Due to the importance of flow determination through rigid vents, a computational fluid dynamics (CFD) model is proposed to verify the capacity of rigid vents in the primary containment of a boiling water reactor (BWR) under different operating conditions (pressure, temperature and compositions of the fluids). The model predicts and provides detailed information on variables such as mass flow and velocity of the venting gases. In the proposed model the primary containment gas is vented to the atmosphere via rigid vents (pipes) from the dry and wet pit. Is assumed that the container is pressurized because is in a defined scenario, and at one point the venting is open and the gas released into the atmosphere. The objective is to characterize the flow and validate the CFD model for the overpressure conditions that occur in an accident such as a LOCA, Sbo, etc. The model is implemented with Ansys-Fluent general-purpose CFD software based on the geometry of the venting ducts of the containment of a BWR. The model is developed three-dimensional and resolves at steady state for compressible flow and includes the effects of the turbulence represented by the Reynolds stress model. The CFD results are compared with the values of a one-dimensional and isentropic model for compressible flow. The relative similarity of results leads to the conclusion that the proposed CFD model can help to predict the rigid venting capacity of the containment of a BWR, however more information is required for full validation of the proposed model. (Author)

  20. Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

    International Nuclear Information System (INIS)

    2000-10-01

    plant components addressed in the reports. This report addresses the metal components of BWR containment systems. The primary ageing mechanisms that may potentially impact the structural capacity, leaktight integrity, or service life of BWR containments are corrosion of metal components and stress corrosion cracking of bellows. Other potential degradation mechanisms include fatigue and mechanical wear. Areas of concern are where surfaces are inaccessible for inspection (e.g areas adjacent to floors, where the containment vessel is embedded in concrete, and locations adjacent to equipment or other structures)

  1. Evaluation of containment failure modes and fission product releases during core meltdown accidents in a BWR with a Mark III containment

    International Nuclear Information System (INIS)

    Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

    1985-01-01

    An assessment is described of potential failure modes and fission product releases for a large number of postulated core meltdown accidents in a BWR with a Mark III containment. For this containment design, the most important failure mode was found to be due to hydrogen related phenomena. A one-dimensional lumped parameter computer code has been developed and used to determine the probability of various hydrogen phenomena for a range of postulated core meltdown sequences. Potential containment loads have been estimated and compared against the containment capacity to determine the probability of containment failure. The fission product release assessment began by using the MARCH/CORRAL system of codes with key input parameters varied over a reasonable range. The parameters relate to primary system retention, re-emission, pool scrubbing, and fission product release in-vessel vs ex-vessel. The final step used more mechanistic calculations based on the system of codes recently developed under sponsorship of the Accident Source Term Program Office, NRC, and compares these predictions with the range of releases calculated in the sensitivity study

  2. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Castillo G, F.

    2015-01-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  3. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  4. Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

    1999-01-01

    In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10 -4 and 10 -6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10 -9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

  5. Mark I BWR pool dynamics: a preliminary investigation into effects of downcomer spacing

    International Nuclear Information System (INIS)

    McCauley, E.W.; Meier, J.K.

    1977-02-01

    A series of experiments were performed to study the effects of downcomer spacing on the growth characteristics of air bubbles. It was found that the momentum of the water thrown up by an air bubble increased with air supply pressure and decreased downcomer spacing. A jet of water formed below the bubble could lead to greater localized loadings on above-pool structures than by the pool swell above the top of the bubble

  6. Full scale BWR containment LOCA response test at the INKA test facility

    International Nuclear Information System (INIS)

    Wagner, Thomas; Leyer, Stephan

    2015-01-01

    KERENA is an innovative boiling water reactor concept with passive safety systems (Generation III+) of AREVA. The reactor is an evolutionary design of operating BWRs (Generation II). In order to verify the functionality and performance of the KERENA safety concept required for the transient and accident management, the test facility “Integral Teststand Karlstein” (INKA) was built at Karlstein (Germany). It is a mock-up of the KERENA boiling water reactor containment, with integrated pressure suppression system. The complete chain of passive safety components is available. The passive components and the levels are represented in full scale. The volume scaling of the containment compartments is approximately 1:24. The reactor pressure vessel (RPV) is simulated via the steam accumulator of the Karlstein Large Valve Test Facility. This vessel provides an energy storage capacity of approximately 1/6 of the KERENA RPV and is supplied by a Benson boiler with a thermal power of 22 MW. With respect to the available power supply, the containment- and system-sizing of the facility is by far the largest one of its kind worldwide. From 2009 to 2012, several single component tests were conducted (Emergency Condenser, Containment Cooling Condenser, Core Flooding System etc.). On March 21st, 2013, the worldwide first large-scale only passively managed integral accident test of a boiling water reactor was simulated at INKA. The integral test measured the combined response of the KERENA passive safety systems to the postulated initiating event was the “Main Steam Line Break” (MSLB) inside the Containment with decay heat simulation. The results of the performed integral test (MSLB) showed that the passive safety systems alone are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them as response to an anticipated accident scenario

  7. The moisture conditions of nuclear reactor concrete containment walls - an example for a BWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, L.O.; Johansson, P. [Lund Institute of Technology, Laboratory of Building Materials, PO Box 118, 221 00 Lund (Sweden)

    2006-07-01

    A method is presented on how to quantify the moisture conditions of nuclear concrete containment walls. The method is based on first quantifying the boundary conditions at the outer and inner surfaces and then describing the moisture fixation and moisture transport within the concrete wall. The temperature and humidity conditions of the outdoor air and of the air close to the wall surfaces are monitored for a period of time and the vapour contents in the different points are compared. From the differences between the vapour contents the sources of moisture are identified and quantified. The previous and future climatic conditions are then predicted. An example is given for the conditions in the containment walls at Barsebaeck nuclear power plant, where moisture measurements have been performed in situ and on samples taken from the walls. (authors)

  8. BWR Passive Containment Cooling System by condensation-driven natural circulation

    International Nuclear Information System (INIS)

    Vierow, K.M.; Townsend, H.E.; Fitch, J.R.; Andersen, J.G.M.; Alamgir, M.; Schrock, V.E.

    1991-01-01

    A method of long-term decay heat removal which is safe, reliable, and passive has been incorporated into the design of the Simplified Boiling Water Reactor (SBWR). The primary functions of the Passive Containment Cooling System (PCCS) are to remove heat and maintain the containment pressure below allowable levels following a LOCA. A key component of the PCCS is the PCC condenser unit (PCC). By natural circulation, a steam-nitrogen mixture flows into the PCC heat exchanger, condensate drains to the reactor pressure vessel (RPV), and noncondensables are vented to the suppression chamber (S/C). This analysis focuses on three significant thermal-hydraulic phenomena which occur in the system. Specifically, steam condensation in the presence of a noncondensable, the PCC noncondensable venting and the natural circulation are discussed. Results of TRACG simulations are presented which show that the PCCS performs its intended functions. (author)

  9. On the failure modes of alternative containment designs following postulated core meltdown

    International Nuclear Information System (INIS)

    Chan, C.K.; Knee, H.E.; Okrent, D.

    1977-01-01

    The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR respresentative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best esimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded. (Auth.)

  10. ZOCO-6, Temperature Transients in BWR and PWR Containment During LOCA

    International Nuclear Information System (INIS)

    Brosche, D.

    1974-01-01

    1 - Nature of physical problem solved: ZOCO-VI is a multi-node computer code which calculates the time and space dependent pressure distribution in containments of water-cooled nuclear power reactors following a loss-of-coolant accident. It should be used primarily for the short term behaviour but can also be used for the long term behaviour. 2 - Method of solution: Using the equations of the mass balances for steam, water and air, and the energy and volume balances, for each pressure node, four or six (non-thermodynamic equilibrium) ordinary non-linear differential equations for the time dependent variations of the temperature and the masses of steam, water and air are obtained. The solution of the system of differential equations will be performed using the integration and iteration procedure DIFSYS. 3 - Restrictions on the complexity of the problem: The quantity of the pressure nodes is only limited by the computer capacity and the computing time

  11. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  12. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    International Nuclear Information System (INIS)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed

  13. An assessment of structural response of condensation pool columns in a BWR/MARK II containment to loads resulting from steam explosions

    International Nuclear Information System (INIS)

    Frid, W.

    1989-01-01

    The objective is to estimate the amount of molten core debris participating in a postulated propagating large-scale steam explosion that could threaten the integrity of the condensation pool columns in a BWR/MARK II containment. This objective was achieved by examination of the structural response of the columns to shock wave loadings and comparison, on the shock wave energy basis, of a propagating steam explosion to a detonation of TNT. In this connection the fraction of the steam explosion energy which appears in the form of a pressure shock wave was estimated. (orig.)

  14. Comparison between MARCH-3 and MAAP-3 thermal-hydraulic results for a severe accident in a BWR system with MARK-III containment

    International Nuclear Information System (INIS)

    Barbucci, P.; Guidi, L.; Mariotti, G.

    1988-01-01

    A comparison between results provided by the Source Term Code Package and by the MAAP-3 code for a PWR with full pressure containment was presented. Thereafter the same two methodologies were used to analyse a severe accident sequence in a typical BWR power plant equipped with a General Electric BWR 6 reactor, rated at 2894 MWt, and a MARK-III type containment. As a reference sequence the TQUV was chosen. This sequence is characterized by a transient (T) with loss of feedwater (Q) and loss of all Emergency Core Cooling Systems, both at high pressure (U) and, after the intervention of the Automatic Depressurization System (ADS), at low pressure (V). After the vessel, failure two basic scenarios for the containment response were analysed: in the first one the pedestal is always dry, in the second one it is fully flooded. Typical limestone/common sand and basaltic concrete compositions were considered. In the following sections the obtained results will be shown with the main purpose to point out the different phenomenological models of the two codes rather than to look for the true plant response to such a severe accident. After the presentation of the most important physical models and of the main assumptions for the analyses (sects. 2 and 4), the comparison will be performed for the in-vessel phase, in section 3, and for the ex-vessel phase, in section 5

  15. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  16. Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

    International Nuclear Information System (INIS)

    McCauley, E.W.; Pitts, J.H.

    1977-01-01

    The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's 1 / 5 -scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized

  17. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  18. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  19. Decommissioning of the ICI TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Parry, D.R.; England, M.R.; Ward, A.; Green, D.

    2000-01-01

    This paper considers the fuel removal, transportation and subsequent decommissioning of the ICI TRIGA Mark I Reactor at Billingham, UK. BNFL Waste Management and Decommissioning carried out this work on behalf of ICI. The decommissioning methodology was considered in the four stages to be described, namely Preparatory Works, Reactor Defueling, Intermediate Level Waste Removal and Low Level Waste Removal. This paper describes the principal methodologies involved in the defueling of the reactor and subsequent decommissioning operations, highlighting in particular the design and safety case methodologies used in order to achieve a solution which was completed without incident or accident and resulted in a cumulative radiation dose to personnel of only 1.57 mSv. (author)

  20. Effects of torus wall flexibility on forces in the Mark I Boiling Water Reactor Pressure Suppression System. Part I

    International Nuclear Information System (INIS)

    Martin, R.W.; McCauley, E.W.

    1977-09-01

    The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems

  1. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde

    International Nuclear Information System (INIS)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C.

    2014-10-01

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  2. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2014-10-01

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm 2 ) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm 2 ), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  3. An assessment of BWR [boiling water reactor] Mark III containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    Schroeder, J.A.; Pafford, D.J.; Kelly, D.L.; Jones, K.R.; Dallman, F.J.

    1991-01-01

    This report describes risk-significant challenges posed to Mark III containment systems by severe accidents as identified for Grand Gulf. Design similarities and differences between the Mark III plants that are important to containment performance are summarized. The accident sequences responsible for the challenges and the postulated containment failure modes associated with each challenge are identified and described. Improvements are discussed that have the potential either to prevent or delay containment failure, or to mitigate the offsite consequences of a fission product release. For each of these potential improvements, a qualitative analysis is provided. A limited quantitative risk analysis is provided for selected potential improvements. 21 refs., 5 figs., 46 tabs

  4. Performance behavior of the passive containment cooling system of a natural circulation BWR during postulated accident condition

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Nayak, A.K.; Jain, Vikas; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Passive systems are playing prominent role in the development of innovative nuclear reactor systems due to their simplicity, enhanced safety, reliability and economy. These systems are being considered for normal operation as well as accidental conditions of reactor following a postulated accident scenario to preclude the scenarios arising out of failure of active systems as well as to minimize the operator intervention. Indian innovative reactor AHWR being designed for thorium utilization employs various passive safety concepts. As containment is the ultimate barrier to the release of radioactivity, passive concepts are being employed in BWRs for minimize peak containment pressure in the containment during a postulated accident condition like LOCA. The concept of passive containment cooling system (PCCS) in the AHWR comprises of inclined tube heat exchangers located underneath an elevated pool that removes the heat from the steam-air atmosphere of containment following a LOCA by natural circulation of water inside the tubes. The steam condenses on the external surface of tubes of PCCS in addition to the wall of the containment which in turn depressurizes the containment. This paper deals with the performance assessment of PCCS of AHWR during a postulated design basis LOCA by using the best estimate code RELAP5/Mod3.2. (author)

  5. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  6. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1991-11-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

  7. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    International Nuclear Information System (INIS)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

  8. EIS pitting temperature determination of A182 nickel based alloy in simulated BWR environment containing dilute seawater

    International Nuclear Information System (INIS)

    Lavigne, Olivier; Shoji, Tetsuo; Takeda, Yoichi

    2014-01-01

    Graphical abstract: - Highlights: • Stable pitting events in function of the temperature are monitored by electrochemical impedance spectroscopy. • The pitting temperature for the nickel based alloy A182 in solution containing 450 ppm Cl − is defined as above 160 °C. • The presented method can be applied for others passive alloys as stainless steel in solution containing aggressive anions. - Abstract: A method based on electrochemical impedance spectroscopy (EIS) measurements to monitor the pitting temperature of passive alloys in a given media is developed in this communication. The pitting corrosion behavior of the nickel based alloy 182 in water containing 450 ppm by weight of chloride is presented in this study. The analysis of the EIS fit parameters from the proposed equivalent electrical circuit allows to determine the temperature from which stable pitting event occurs at open circuit potential. For the A182 sample this temperature is measured above 160 °C

  9. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  10. Development and operational experiences of an automated remote inspection system for interior of primary containment vessel of a BWR

    International Nuclear Information System (INIS)

    Ozaki, N.; Chikara, S.; Fumio, T.; Katsuhiro, M.; Katsutoshi, S.; Ken-Ichiro, S.; Masaaki, F.; Masayoshi, S.

    1983-01-01

    A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests

  11. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  12. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  13. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  14. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    International Nuclear Information System (INIS)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

  15. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  16. Vent clearing during a simulated loss-of-coolant accident in a Mark I boiling-water reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    In this test series, drywell pressurization rate, drywell overpressure, downcomer submergence, and overall vent system loss coefficient were varied to quantify the primary load sensitivities in the pressure suppression system. Extensive tests were conducted on a unique three-dimensional 1/5 scale model of the pressure suppression system a MARK-I BWR. They were focused on the initial or air cleaning phase of a hypothetical loss of coolant accident. As a result of the complete measurement system employed including multiple high speed cameras, the logical interrelationship between measured forces, measured pressures, and the hydrodynamic phenomena observed in high speed photographic pictures were established. The quantitative values from the 1/5 scale experiments can be applied to full scale plants using established scaling laws. (author)

  17. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  18. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1978-01-01

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

  19. Measuring temperature coefficient of TRIGA MARK I reactor by noise analysis

    International Nuclear Information System (INIS)

    Soares, P.A.

    1975-01-01

    The transfer function of TRIGA MARK I Reactor is measured at power zero (5w) and power 118Kw, in the frequency range of 0.02 to 0.5 rd/s. The method of intercorrelation between a pseudostochasticbinary signal is used. A simple dynamic model of the reactor is developed and the coefficient of temperature is estimated [pt

  20. Insights for aging management of light water reactor components: Metal containments

    International Nuclear Information System (INIS)

    Shah, V.N.; Sinha, U.P.; Smith, S.K.

    1994-03-01

    This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel

  1. Timing criteria for supplemental BWR emergency response equipment

    International Nuclear Information System (INIS)

    Bickel, John H.

    2015-01-01

    The Great Tohuku Earthquake and subsequent Tsunami represented a double failure event which destroyed offsite power connections to Fukushima-Daiichi site and then destroyed on-site electrical systems needed to run decay heat removal systems. The accident could have been mitigated had there been supplemental portable battery chargers, supplemental pumps, and in-place piping connections to provide alternate decay heat removal. In response to this event in the USA, two national response centers, one in Memphis, Tennessee, and another in Phoenix, Arizona, will begin operation. They will be able to dispatch supplemental emergency response equipment to any nuclear plant in the U.S. within 24 hours. In order to define requirements for supplemental nuclear power plant emergency response equipment maintained onsite vs. in a regional support center it is necessary to confirm: (a) the earliest time such equipment might be needed depending on the specific scenario, (b) the nominal time to move the equipment from a storage location either on-site or within the region of a nuclear power plant, and (c) the time required to connect in the supplemental equipment to use it. This paper describes an evaluation process for a BWR-4 with a Mark I Containment starting with: (a) severe accident simulation to define best estimate times available for recovery based on the specific scenario, (b) identify the key supplemental response equipment needed at specific times to accomplish recovery of key safety functions, and (c) evaluate what types of equipment should be warehoused on-site vs. in regional response centers. (authors)

  2. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  3. Range of the radiation monitor for the rigid vent of primary containment during normal and emergency operation for a BWR-5 in Laguna Verde; Rango del monitor de radiacion para el venteo rigido de la contencion primaria durante operacion normal y emergencia para un reactor BWR-5 en Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F.; Pozos S, A. M.; Cabrera U, S.; Mata A, J. A.; Sandoval V, S.; Ovando C, R.; Vargas A, A.; Gallardo R, I.; Cruz G, M.; Amador C, C., E-mail: francisco.tijerina@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The earthquake followed by a tsunami, happened in March, 2011 in the coasts of oriental Japan, caused damages in the nuclear power plants 1 at 4 of Fukushima Daiichi leading to damage of the fuel in three of the reactors and to the radiation liberation to the exterior. As consequence of those events, the regulations requires that the power plants with Primary Containment type Mark I and II evaluate to have a system of rigid vent with a monitoring equipment of radiation effluents. The present work covers the rigid vent of diameter 12 of the Primary Containment, type Mark-II, of nuclear power plant of Laguna Verde in conditions of severe accident and normal operation, low regime of Extended Power Up rate (EPU - 2317 MWt), using the codes MAAP3B, MICROSHILED 5.05 and the Bardach Black Boxes methodology. As a result the measurement range of the radiation monitor that is required for monitoring the gassy liberation to the atmosphere was determined. The conclusion is that the superior limit of the range of the radiation meter during a Severe Accident is of 8.55 E + 05 R/h (8.55 E + 08 m R/h) and the superior limit in normal operation of 1.412 E-11 at 2.540 E-7 R/h (1.412 E-14 at 2.540 E-10 m R/h). (Author)

  4. PSEPLOT: a controller for plotting data from the Mark I Boiling Water Reactor Pressure Suppression Experiment

    International Nuclear Information System (INIS)

    Holman, G.S.

    1978-01-01

    PSEPLOT is a computer routine that was developed for the Lawrence Livermore Laboratory Octopus computer system to generate several thousand plots of engineering data in a consistent format for referencing and comparison. The time-dependent engineering data were recorded during each of 25 tests of the Mark I Pressure Suppression Experiment (PSE). Although PSEPLOT is restricted to PSE, its concept is applicable to any similar data management task

  5. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5; Simulacion de un escenario de perdida total de potencia externa e interna (SBO) para distintas presiones de venteo de la contencion de un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm{sup 2}) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm{sup 2}), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  6. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  7. Analysis of containment venting for the Peach Bottom Atomic Power Station

    International Nuclear Information System (INIS)

    Hanson, D.J.; Wright, R.E.; Jenkins, J.P.

    1986-01-01

    The effectiveness of containment venting as a means of preventing or mitigating the consequences of severe accidents was evaluated for Peach Bottom Atomic Power Station Units 2 and 3 (BWR-4s with Mark I containments). Results from this evaluation indicate that the effectiveness of venting in preventing containment failure is highly dependent on the severe accident sequence. Containment venting can be effective for several classes of sequences, including loss-of-coolant accidents with breaks in the containment and transients with a failure of containment heat removal. However, based on draft procedures and equipment in place at the time of the evaluation, containment venting has limited potential for further reducing the risk associated with several sequences currently identified as significant contributors to risk. Means of improving the potential for risk reduction were identified, but their influence on risk was not analyzed

  8. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  9. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  10. Instrumenting a pressure suppression experiment for a Mark I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1978-01-01

    A 1 / 5 -scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system

  11. Neutron flux determination at the IPR-R1 Triga Mark I neutron beam extractor

    International Nuclear Information System (INIS)

    Zangirolami, Dante Marco; Maretti Junior, Fausto; Ferreira, Andrea Vidal

    2009-01-01

    The IPR-R1 Triga Mark I Reactor located at the CDTN/CNEN, Belo Horizonte, Brazil, has been operating since November of 1960. In this work, measurements of thermal and epithermal neutron flux along the IPR-R1 neutron beam extractor were performed by neutron activation of reference materials using the two foils method. The obtained results were compared with results from two previous works: an experimental measurement done in a previous reactor core configuration and a numerical work made by Monte Carlo simulation using the actual reactor core configuration. The main purpose of this work is to update the measured data to the actual reactor core configuration. (author)

  12. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  13. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1993-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments

  14. Analysis of containment venting at the Peach Bottom atomic power station

    International Nuclear Information System (INIS)

    Hanson, D.J.; Blackman, H.S.; Nelson, W.R.; Wright, R.E.; Leonard, M.T.; DiSalvo, R.

    1988-01-01

    An analysis of the extent to which containment venting would be effective in preventing or mitigating the consequences of severe accidents has been completed for the Peach Bottom Units 2 and 3 (BWR-4s with Mark I containments). The analysis indicates that the effectiveness of venting in preventing containment overpressurization highly depends on the sequence of the severe accident. Containment venting can be effective for several classes of sequences, including transients with failure of long-term decay heat removal and loss-of-coolant accidents with breaks inside the containment. However, based on draft procedures and equipment in place at the time of the evaluation, containment venting has limited potential for further reducing the risk associated with three severe accident sequences currently identified as important risk contributors at Peach Bottom. Means of improving the potential for risk reduction is identified, but their influence on risk is not analyzed. (orig./HP)

  15. Analysis of containment venting at the Peach Bottom Atomic Power Station

    International Nuclear Information System (INIS)

    Hanson, D.J.; Blackman, H.S.; Nelson, W.R.; Wright, R.E.; Leonard, M.T.; DiSalvo, R.

    1986-01-01

    An analysis of the extent to which containment venting would be effective in preventing or mitigating the consequences of severe accidents has been completed for the Peach Bottom Atomic Power Station, Units 2 and 3 (BWR-4s with Mark I containments). The analysis indicates that the effectiveness of venting in preventing containment overpressurization highly depends on the sequence of the severe accident. Containment venting can be effective for several classes of sequences, including transients with failure of long-term decay heat removal and loss-of-coolant accidents with breaks inside the containment. However, based on draft procedures and equipment in place at the time of the evaluation, containment venting has limited potential for further reducing the risk associated with three severe accident sequences currently identified as important risk contributors at Peach Bottom. Means of improving the potential for risk reduction is identified, but their influence on risk is not analyzed

  16. Assessment of corrosion and fatigue damage to light water reactor metal containments

    International Nuclear Information System (INIS)

    Sinha, U.P.; Shah, V.N.; Smith, S.K.

    1991-01-01

    This paper presents a generic procedure for estimating aging damage, evaluating structural integrity, and identifying mitigation activities for safe operation of boiling water reactor (BWR) Mark I metal containments and ice-condenser type pressurized water reactor (PWR) cylindrical metal containments. The mechanisms of concern that can cause aging damage to these two types of containments are corrosion and fatigue. Assessment of fatigue damage to bellows is also described. Assessment of corrosion and fatigue damage described in this paper include: containment design features that are relevant to aging assessment, several corrosion and fatigue mechanisms, inspection of corrosion and fatigue damage, and mitigation of damage caused by these mechanisms. In addition, synergistic interaction between corrosion and fatigue is considered. Possible actions for mitigating aging include enhanced inspection methods, maintenance activities based on operating experience, and supplementary surveillance programs. Field experience related to aging of metal containments is reviewed. Finally, conclusions and recommendations are presented

  17. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  18. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  19. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blotcky, A J; Arsenault, L J [General Medical Research, Veterans Administration Hospital, Omaha (United States)

    1974-07-01

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  20. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenault, L.J.

    1974-01-01

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  1. TRIGASIM: A computer program to simulate a TRIGA Mark I Reactor

    International Nuclear Information System (INIS)

    Ruby, Lawrence

    1992-01-01

    A Fortran-77 computer program has been written which simulates the operation of a TRIGA Mark I Reactor. The 'operator' has options at 1-second intervals, of raising rods, lowering rods, maintaining rods steady, dropping a rod, or scramming the reactor. Results are printed to the screen, and to 2 output files - a tabular record and a logarithmic plot of the power. The Point Kinetic Equations are programmed with 6 delayed groups, quasi-static power feedback, and forward differencing. A pulsing option is available, with simulation which employs the Fuchs Model. A pulse-tail model has been devised to simulate behavior for a few minutes following a pulse. Both graphic and tabular output are also available for the pulses. (author)

  2. Estimation of sesqui-carbide fraction for MARK-I fuel

    International Nuclear Information System (INIS)

    Vana Varamban, S.; Ananthasivan, K.

    2016-01-01

    Sesqui-carbide content of FBTR bi-phasic mixed carbide is specified as 5-20 wt.%. For each batch of fuel production, the sesqui-carbide (M2C3) content is being determined by a K-ratio method using XRD information. There is a need to evolve an alternate method for qualitative determination of M2C3 content for a fabricated FBTR fuel pellet. Two independent approaches resulted in a correlation between overall carbon content and the M2C3 phase fraction. The thermodynamic calculations agree well with the stoichiometric correlation between the overall carbon content and the M2C3 phase fraction in FBTR MARK I fuel

  3. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  4. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  5. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  6. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  7. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  8. Dry reloading and packaging of spent fuel at TRIGA MARK I reactor of Medical University Hanover (MHH), Germany

    International Nuclear Information System (INIS)

    Haferkamp, D.

    2008-01-01

    which is protected against contamination spreading and provided with a conventional driven by a floor conveyor trolley. 4. Mobile Reloading Station placed directly on the shipping cask, containing adapter to each type of shipping cask, shielded sliding plates and provided with a loading plate for positioning of six packs into the shipping cask. 5. Auxiliary systems including: - cask lid lifting device; - video monitoring device for the shipping cask lids and loads; - air film gliders for transport of shipping casks. Finally one concludes: the dry reloading station for spent fuel packaging into the shipping casks has successfully operated for TRIGA MARK I reactor at Medical University Hanover as well as at the VKTA - Research Reactor in Rossendorf, Germany. It is available as a proven technique for the dry reloading of spent fuels into the shipping casks. (author)

  9. Containment loadings due to hydrogen burning in LWR core meltdown accidents

    International Nuclear Information System (INIS)

    Cybulskis, P.

    1981-01-01

    The potential pressure loadings due to hydrogen burning under conditions representative of meltdown accident conditions are examined for a variety of PWR and BWR containment designs. For the PWR, the large dry, ice condenser, as well as subatmospheric containments are considered. For the BWR, MARK I, II, and III pressure suppression containments are evaluated. The key factors considered are: free volume, design pressure, extend to hydrogen generation, and the flammability of the atmosphere under a range of accident conditions. The potential for and the possible implications of hydrogen detonation are also considered. The results of these analyses show that the accumulation and rapid burning of the quantities of hydrogen that would be generated during core meltdown accidents will lead to pressures above design levels in all of the containments considered. As would be expected, containments characterized by small volumes and/or low design pressures are the most vulnerable to damage due to hydrogen burning. Large volume, high pressure designs may also be threatened but offer significantly more potential for accomodating hydrogen burns. The attainment of detonable hydrogen mixtures is made easier by smaller containment volumes. Detonable mixtures are also possible in the larger volume containments, but imply the accumulation of hydrogen for long periods of time without prior ignition. Hydrogen detonations, if they occur, would probably challenge the integrity of any of the containments considered. (orig.)

  10. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  11. IPR-RI TRIGA MARK I reactor and the neutron activation analysis at CDTN/CNEN

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Kastner, Geraldo F.; Amaral, Angela M.; Souza, Wagner de; Maretti, Fausto Junior; Leal, Alexandre S.

    2008-01-01

    The IPR-R1 TRIGA Mark I research reactor started up in 1960. It is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. Join to the reactor, the Laboratory for Neutron Activation Analysis has been developing its activities since 1960. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for relevant percentage of CDTN's analysis demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays several elements - Ag, Al, Au, As, Ba, Br, Ca, Cd, Ce, Cl, Co, Cr, Cs, Cu, Dy, Eu, Fe, Ga, Hf, Hg, Ho, K, La, Mg, Mn, Mo, Na, Nd, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, Ti, U, V, W, Yb, Zn and Zr - are determined in several matrices and range of concentrations. In Brazil, CDTN is the only Institute that fully masters the instrumental neutron activation analysis k0-method determining short, medium and long half-life radionuclides using its own nuclear reactor. The good performance of the reactor is pointed out in a table with experimental and certified values for Certified Reference Materials. (authors)

  12. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  13. The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

    International Nuclear Information System (INIS)

    Maretti, Fausto Junior; Sette Camara, Luiz Otavio I.; Oliveira, Paulo Fernando

    2008-01-01

    The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

  14. Source term attenuation by water in the Mark I boiling water reactor drywell

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A. [Sandia National Labs., Albuquerque, NM (United States)

    1993-09-01

    Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm{sup 3} H{sub 2}O/cm{sup 2}-s and decontamination factors of 1.1, 2, 3.3, 10, 100, and 1000.

  15. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    International Nuclear Information System (INIS)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

  16. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR's as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs

  17. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  18. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  19. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  20. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  1. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

    ... of Operation Under Severe Accident Conditions (Effective Immediately) I. The Licensees identified in... radioactive materials if the venting systems were used during severe accident conditions. The NRC staff..., 2012). Option 2 in SECY-12-0157 was to modify EA-12-050 to require severe accident capable vents (i.e...

  2. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  3. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  4. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  5. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  6. Metallurgical factors that contribute to cracking in BWR piping

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

  7. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  8. Phase II investigation of the response of columns to short duration loadings: scaling report. Mark I Program, Task 3.2.1. Teledyne report TR-2488(e)

    International Nuclear Information System (INIS)

    1978-03-01

    Phase II of a two-phase column testing program which will determine short-term load capacities for the Mark I containment columns is now under way. The scaling report is submitted to outline the scaling procedures and also to present the scaled quantities such as geometry, load eccentricity, and load time duration. Dimensional analysis and similarity techniques were applied to the experimental models to obtain a set of twelve scaling laws. These scaling laws provide a rational basis for modeling the columns. The scaling laws give nondimensional relationships which are common to both the model and prototype. Parameters which are known to be important in column buckling and which were found to be important during Phase I testing were used as the independent variables in the scaling laws. Column specimens for Phase II testing have been designed to be representative of the prototype columns. Combined bending and compression are characteristic of the Mark I columns with the maximum moment applied at the column-ring girder junction. The experimental columns reflect these characteristics, even though special fixtures and equipment will be required to perform the tests. The column specimens include the scaled rotational stiffness of the ring girder as well as the boundary condition of the pinned or sliding end. To properly incorporate the effects of residual stress, the models will be constructed from ''off-the-shelf', carbon steel pipe and rolled sections, and built-up sections will be fabricated from standard plate. The column geometries scale almost exactly, even though ''replica'' modeling was not used

  9. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  10. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  11. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  12. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  13. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

  14. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  15. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1994-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

  16. Identification and assessment of containment and release management strategies

    International Nuclear Information System (INIS)

    Lehner, J.R.; Lin, C.C.; Neogy, P.

    1990-01-01

    Brookhaven National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize releases during a severe accident. The objective is to make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident (i.e., after vessel breach) and sizeable uncertainties exist regarding some of the phenomena involved. A majority of the strategies identified go well beyond existing procedures and often depend on the specific containment type. Strategies for all of the five different containments used in the U.S. are being considered: BWR Mark I, Mark II, and Mark III, as well as PWR ice condenser and large dry containments. Accident management strategies related to the in-vessel phase of a severe core melt accident are being dealt with under another NRC program. For each containment type the most likely challenges are identified and existing emergency guidelines and procedures are reviewed as to how they address these challenges

  17. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  18. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  19. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  20. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  1. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  2. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  3. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  4. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  5. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  6. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  7. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  8. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  9. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  10. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  11. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

  12. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    Tavera D, L.; Camacho L, M.E.

    1991-01-01

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  13. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  14. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  15. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L; Camacho L, M E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  16. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  17. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  18. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  19. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  20. Identification and assessment of containment and release management strategies

    International Nuclear Information System (INIS)

    Lehner, J.R.; Lin, C.C.; Neogy, P.

    1993-01-01

    Brookhaven National Laboratory, under the auspices of the U.S. Nuclear Regulatory Commission, is investigating accident management strategies which could help preserve containment integrity or minimize the release of radioactivity during a severe accident in a nuclear reactor. The objective is to make use of existing plant systems and equipment in innovative ways to reduce the likelihood of containment failure or to mitigate the release of fission products to the environment if failure cannot be prevented. Many of these strategies would be implemented during the later stages of a severe accident, i.e. after the molten core penetrates the reactor vessel. Significant uncertainties exist regarding some of the phenomena involved with this phase of a severe accident. The identification and assessment process for containment and release strategies is described, and some insights derived from its application to a BWR Mark I plant are presented. A station blackout accident for this kind of plant is considered. The challenges encountered are identified and existing emergency guidelines are reviewed, where needed and when possible, new strategies are devised. The feasibility and effectiveness of these new strategies are assessed, making due allowances for the complicated phenomena and associated uncertainties involved. Both beneficial and adverse effects of the suggested strategies are considered. (orig.)

  1. The past and the future in the forty years of the IPR-R1 TRIGA MARK I reactor operation

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto

    2008-01-01

    Full text: The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. During these years a lot of irradiations, analysis , MSc and PhD thesis, training courses and isotopes production take place at the reactor. This paper describes the improvements made, the results obtained during the past 40 years, type of works realized, isotopes produced, the neutron activation analysis and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (authors)

  2. Neutron activation analysis at CDTN/CNEN using the IPR-R1 Triga Mark I reactor

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Maretti Junior, Fausto; Kastner, Geraldo Frederico; Amaral, Angela Maria; Souza, Wagner de

    2009-01-01

    This paper describes in summary the activities developed by the Laboratory for Neutron Activation Analysis since the starting up of the IPR-R1 TRIGA Mark I research reactor in 1960. This Laboratory is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for significant percentage of CDTN's analytical demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays the neutron activation analysis is responsible for 70% of the analytical demand and the k 0 - Instrumental method for 80% of this demand answering clients' request and researches. In Brazil, CDTN is the only Institute that fully masters the Instrumental Neutron Activation Analysis k 0 -method using its own nuclear reactor. (author)

  3. Analysis of BWR/Mark III drywell failure during degraded core accidents

    International Nuclear Information System (INIS)

    Yang, J.W.

    1983-01-01

    The potential for a hydrogen detonation due to the accumulation of a large amount of hydrogen in the drywell region of a BWR Mark III containment is analyzed. Loss of integrity of the drywell wall causes a complete bypass of the suppression pool and leads to pressurization of the containment building. However, the predicted peak containment pressure does not exceed the estimates of containment failure pressure

  4. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Ott, L.J.

    1997-01-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  5. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  6. Corrosion issues in the BWR and their mitigation for plant life extension

    International Nuclear Information System (INIS)

    Gordon, B.M.

    1988-01-01

    Corrosion is a major service life limiting mechanism for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). For the BWR, stress corrosion cracking of piping has been the major source of concern where extensive research has led to a number of qualified remedies and currently > 90% of susceptible welds have been mitigated or replaced. Stress corrosion cracking of reactor internals due to the interaction of irradiation, as discussed elsewhere in this conference, is also a possible life limiting phenomenon. This paper focusses on two corrosion phenomena in the BWR which have only recently been identified as impacting the universal goal of BWR life extension: the general corrosion of containment structures and the erosion-corrosion of carbon steel piping

  7. Recent insights from severe accident research and implications for containment criteria for advanced LWRs

    International Nuclear Information System (INIS)

    Speis, T.P.; King, T.L.; Eltawila, F.

    1992-01-01

    The Severe Accident Research Program (SARP) was begun after the TMI-2 accident in March, 1979. The rule for dealing with the generation of large quantity of hydrogen in BWRs and Ice Condenser PWRs was promulgated by the Nuclear Regulatory Commission (NRC). The NRC issued severe Accident Policy Statement in 1985, and the revised SARP in 1989. In this paper, the current understanding of the more important phenomena and the associated mechanical and thermal loads to the containment is described, and the on-going works are summarized. The containment loadings in severe accidents are listed, and direct containment heating and the liner failure in BWR Mark I are added. A great deal of informations obtained on the early phase of melt progression are shown. The current understanding of the severe accident phenomena related to the containment and the on-going related research efforts are discussed more in detail. Fuel-coolant interaction including alpha-mode containment failure, direct containment heating, hydrogen deflagration and detonation, core-concrete interaction and debris coolability are described. (K.I.)

  8. Recently-developed neutron activation analysis techniques utilizing the University of California at Irvine TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Guinn, V.P.; Chambless, D.; Cortes T, E.; DeLancey, K.; Garzonov, E.; Miller, D.A.; Miller, G.E.; Purcell, M.A.

    1976-01-01

    The University of California at Irvine (UCI) 250 kW TRIGA Mark I reactor is used extensively for neutron activation analysis (NAA) studies. These particularly include basic technique studies and application studies in the fields of environmental pollution, crime investigation, archaeology, oceanography, and geochemistry. In recent NAA studies at UCI, a number of techniques have been developed which considerably improve the usefulness of such a research reactor for NAA work, and which should be of interest and use to others. Six of these techniques will be described in further detail in the full paper. They are as follows: development and use of (1) an automated high-precision rapid transfer system for instrumental NAA measurements with induced activities having half lives as short as 0.5 second, (2) an automated measurement system and computer program for making accurate dead-time corrections under conditions where the Ge(Li) spectrometer deadtime is changing rapidly during the counting period, (3) a technique to minimize the loss of mercury from samples during reactor irradiation via the use of dry-ice-packed, vented TRIGA rotary rack tubes, (4) a technique for compacting powdered samples, by pre-irradiation treatment with a solution of paraffin in carbon disulfide, to provide reproducible irradiation and counting geometries, (5) a method utilizing hydrated antimony pentoxide (HAP) as a pre-irradiation treatment material for removal of sodium from aqueous and wet-ashed samples, and (6) a computerized system for predicting in advance of activation, from approximate known elemental compositions, the total counting rate, deadtime, spectrum shape, principal photopeaks, and approximate actual lower limits of instrumental NAA detection of designated elements for any selected irradiation and decay times. (author)

  9. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Karjunen, T [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T; Piippo, J; Sirkiae, P [VTT Manufacturing Technology (Finland)

    1996-12-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

  10. Radiation buildup and control in BWR recirculation piping

    International Nuclear Information System (INIS)

    Meyer, W.; Wood, R.M.; Rao, T.V.; Vook, R.W.

    1987-01-01

    Boiling water nuclear reactors (BWRs) employ stainless steel (Types 304 or 316 NG) pipes in which high-purity water at temperatures of ∼ 275 0 C are circulated. Various components of the system, such as valves and bearings, often contain hard facing metal alloys such as Stellite-6. These components, along with the stainless steel tubing and feedwater, serve as sources of 59 Co. This cobalt, along with other soluble and insoluble impurities, is carried along with the circulating water to the reactor core where it is converted to radioactive 60 Co. After reentering the circulating water, the 60 Co can be incorporated into a complex corrosion layer in the form of CoCr 2 O 4 and/or CoFe 2 O 4 . The presence of even small amounts of 60 Co on the walls of BWR cooling systems is the dominant contributor to inplant radiation levels. Thus BWR owners and their agents are expending significant time and resources in efforts to reduce both the rate and amount of 60 Co buildup. The object of this research is twofold: (a) to form a thin diffusion barrier against the outward migration of cobalt from a cobalt-containing surface and (b) to prevent the growth of a 60 Co-containing corrosion film. The latter goal was the more important since most of the radioactive cobalt will originate from sources other than the stainless steel piping itself

  11. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  12. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  13. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  14. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  15. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  16. Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1978-01-01

    The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

  17. Study on alternatives of inertisation of nuclear power plant containment

    International Nuclear Information System (INIS)

    Baron, J.H.; Zarate, S.M.

    1998-01-01

    In the course of a severe accident in a nuclear power plant, the hydrogen generation and other flammable gases, during the core degradation phase and the interaction corium-concrete, could produce the failure of the containment by overpressure of by combustion. According to the analysis of the potential effects of hydrogen evolution, following accidents inside the containment trough a Defense-in depth principle, which attempts to assure that the containment must not fail catastrophically, two techniques have been evaluated: a: Inertisation pre-accident and b: Inertisation post-accident. The technique of inertisation pre-accident consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ) during the normal operation. The inertisation post-accident in combination with early venting system consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ), immediately after the beginning of the accident, while the radioactivity is still negligible inside the containment. A system of inertisation pre-accident with nitrogen is used on BWR Mark I and Mark II. Investigations on the inertisation post-accident of the containment atmosphere during severe accidents have been carried out with different objectives from principles of the decade of 1980. Studies concerning hydrogen problem for the nuclear power plants Atucha I and CAREM-25 have permitted to know that the hydrogen generation during an accidental sequence with core degradation, would result important, being able to arrive to form explosive mixtures. In the present work, the applicability of the techniques of inertisation is analyzed for the containment of the Atucha I and CAREM-25, considering the particular design characteristics of these plants. (author) [es

  18. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

    2015-04-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

  19. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

    Castillo, Rogelio; Ramírez, J. Ramón; Alonso, Gustavo; Ortiz-Villafuerte, Javier

    2015-01-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred

  20. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  1. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  2. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  3. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  4. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  5. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  6. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350 0 F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage

  7. Moderator temperature coefficient in BWR core

    International Nuclear Information System (INIS)

    Naito, Yoshitaka

    1977-01-01

    Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

  8. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  9. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  10. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  11. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  12. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  13. Structural test and analysis of a model of a BWR suppression chamber support in the plastic regime

    International Nuclear Information System (INIS)

    Blumer, U.R.; Klaeui, E.; Bosshard, E.P.

    1991-01-01

    A BWR Mark I suppression pool support has been analysed and tested in the laboratory. The aim was the demonstration of the acceptability of hypothetical dynamic loadings resulting from simultaneous steam blowdown through all safety relief valves. The analysis has shown that plastic deformation will locally occur, which is difficult to assess purely theoretical. Therefore tests in reduced scale were performed that show the amount and distribution of plastic flow in the supports. The paper describes the elastic analysis, the theory of the scaling laws for the reduced scale test, the test and its results. It also shows the thermographical method that has been used to determine the plastic material flow in the support structure. (author)

  14. Development and recent trend of design of BWR nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kani, J [Tokyo Shibaura Electric Co. Ltd., Kawasaki, Kanagawa (Japan)

    1977-11-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation.

  15. Development and recent trend of disign of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Kani, Jiro

    1977-01-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

  16. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  17. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (author)

  18. Analytical model for computing transient pressures and forces in the safety/relief valve discharge line. Mark I Containment Program, task number 7.1.2

    International Nuclear Information System (INIS)

    Wheeler, A.J.

    1978-02-01

    An analytical model is described that computes the transient pressures, velocities and forces in the safety/relief valve discharge line immediately after safety/relief valve opening. Equations of motion are defined for the gas-flow and water-flow models. Results are not only verified by comparing them with an earlier version of the model, but also with Quad Cities and Monticello plant data. The model shows reasonable agreement with the earlier model and the plant data

  19. Hydrogen evolution from aluminium in reactor containment spray solutions

    International Nuclear Information System (INIS)

    Karlberg, G.; Sundvall, S.-B.

    1982-01-01

    Three different aluminium alloys were exposed to conditions similar to BWR and PWR containment spray waters at 50, 100 and 150 0 C. BWR deionized water gives corrosion rates of at most 0.05 mm/year and hydrogen concentrations less than 0.1-1%. On the contrary PWR alkaline solutions give very high corrosion rates and hydrogen contents. (Auth.)

  20. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  1. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  2. Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Larry J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate

  3. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  4. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  5. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

  6. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  7. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  8. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  9. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  10. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

  11. IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1996-01-01

    Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

  12. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  13. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  14. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  15. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  16. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  17. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  18. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

  19. Efficient method for simulation of BWR severe accident sequence events before core uncovery

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1984-01-01

    BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are modeled. Valuable insights have been gained into accident sequences. A Fortran version is under development and it will be modified for application to Mark II plants

  20. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document Volume 3, contains the Bases for Sections 3.4--3.10 of the improved STS

  1. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  2. Standard Technical Specifications, General Electric Plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  3. Standard Technical Specifications, General Electric plants, BWR/6

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/6, and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. This document Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  4. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  5. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  6. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  7. Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

    International Nuclear Information System (INIS)

    Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

    2008-01-01

    As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

  8. A simplified spatial model for BWR stability

    International Nuclear Information System (INIS)

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-01-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  9. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  10. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  11. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  12. BWR series pump recirculation system

    International Nuclear Information System (INIS)

    Dillmann, C.W.

    1992-01-01

    This patent describes a recirculation system for driving reactor coolant water contained in an annular downcomer defined between a boiling water reactor vessel and a reactor core spaced radially inwardly therefrom. It comprises a plurality of circumferentially spaced second pumps disposed in the downcomer, each including an inlet for receiving from the downcomer a portion of the coolant water as pump inlet flow, and an outlet for discharging the pump inlet flow pressurized in the second pump as pump outlet flow; and means for increasing pressure of the pump inlet flow at the pump inlet including a first pump disposed in series flow with the second pump for first receiving the pump inlet flow from the downcomer and discharging to the second pump inlet flow pressurized in the first pump

  13. ECP measurements in the BWR-1 water loop relative to water composition changes

    Energy Technology Data Exchange (ETDEWEB)

    Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2010-07-01

    The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

  14. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  15. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

    2010-01-01

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  16. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  17. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  18. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  19. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  20. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  1. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  2. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  3. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

  4. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  5. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  6. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  7. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  8. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

  9. Best-estimate analysis development for BWR systems

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

    1986-01-01

    The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

  10. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  11. BWR stability: history and state-of-the-art

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2014-01-01

    The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

  12. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  13. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  14. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  15. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

  16. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  17. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  18. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  19. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J.M.; Blazquez, J.B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical struct-ure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  20. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  1. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  2. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  3. BWR alloy 182 stress Corrosion Cracking Experience

    International Nuclear Information System (INIS)

    Horn, R.M.; Hickling, J.

    2002-01-01

    Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

  4. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

  5. Simulation of hydrogen deflagration and detonation in a BWR reactor building

    International Nuclear Information System (INIS)

    Manninen, M.; Silde, A.; Lindholm, I.; Huhtanen, R.; Sjoevall, H.

    2002-01-01

    A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm 2 leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis

  6. BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    Anstine, L.D.

    1983-05-01

    This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

  7. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  8. Reference core design Mark-I and -II of the experimental, multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Hirano, Mitsumasa; Aruga, Takeo; Yasukawa, Sigeru

    1977-10-01

    Reactivity worth of the control rods and power distribution in the initial hot-clean core of reference core design Mark-I and -II have been studied. The need for burnable poison was confirmed, because of the limitations in number, diameter and reactivity worth of the control rods due to structures of pressure vessel and fuel element and to safety of the core. While the initial excess reactivity is reduced by use of the burnable poison, the recovery of core reactivity with burnup of the burnable poison requires a complicated withdrawal sequence of the control rods. The radial power gradient in the core is not large, due to orifice control of the coolant helium flow, effectiveness of the reflector in the small core and continuous distribution of burnup in the core by one-batch refuelling scheme. The local peaking factor in unit orifice regions, therefore, is the most important core design. Control of the axial power distribution is necessary to reduce the maximum fuel temperature and the exponential power distribution peaked toward the inlet of the core is most suitable. However, insertion of the control rods from top of the core disturbs the axial power distribution, so this effect must be considered in design of the withdrawal sequence of control rods. Nuclear properties of the core were revealed from results of the study for the initial hot-clean core. (auth.)

  9. Thermal power calibration of the TRIGA Mark I IPR-R1 reactor during the upgrading tests to 250 kW

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Maretti, Fausto Junior; Rezende, Hugo Cesar

    2002-01-01

    This paper presents the results and the methodology used to calibrate the thermal power of the TRIGA MARK I IPR-R1 Reactor in CDTN, Belo Horizonte, Brazil. This calibration was realized during the operation tests carried out to allow the reactor power upgrade from the current 100 kW to 250 kW. The methodology consisted in the measurement of the inlet and outlet temperature and the water flow in the primary cooling loop. The thermal balance together with the thermal losses gave the thermal power. There were made three sequences of tests. The first rising of the thermal power was made with the usual configuration of the core (59 fuel elements). After the changing of the ion chambers position and the control rod and the increase of the number of fuels (63 fuel elements), a new evaluation of the thermal power was accomplished, having been obtained a thermal power of 234 kW, for an indication of 250 kW in the lineal channel. After the return of the core to the initial configuration (59 fuel elements), it took place a new test, getting back the reactor to the power level of 100 kW. (author)

  10. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

    Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

  11. Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

    Directory of Open Access Journals (Sweden)

    Tanaka Ken-ichi

    2016-01-01

    Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

  12. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  13. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  14. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

    2014-05-15

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

  15. Water level measurement uncertainty during BWR instability

    International Nuclear Information System (INIS)

    Torok, R.C.; Derbidge, T.C.; Healzer, J.M.

    1994-01-01

    This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by ppercase RETRAN for an ATWS/instability event.The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument. ((orig.))

  16. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  17. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  18. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  19. Method of operating BWR type power plants

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

  20. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  1. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  2. Primary coolant system of BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

    1997-01-01

    The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

  3. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  4. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  5. Numerical simulation of boron injection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

    2010-02-15

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

  6. Numerical simulation of boron injection in a BWR

    International Nuclear Information System (INIS)

    Tinoco, Hernan; Buchwald, Przemyslaw; Frid, Wiktor

    2010-01-01

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several

  7. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  8. Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

    2003-01-01

    Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

  9. Alternative Zr alloys with irradiation resistant precipitates for high burnup BWR application

    International Nuclear Information System (INIS)

    Garzarolli, F.; Ruhmann, H.; Van Swan, L.

    2002-01-01

    In the core of BWRs, the second-phase particles (SPP) of Zircaloy-2 and Zircaloy-4, the Zr(FeCr) 2 and the Zr 2 (FeNi) phase, release Fe and dissolve. The degree of dissolution depends on initial size and fluence. These SPP, however, are important for the corrosion behavior of Zircaloy. Zircaloy shows an increase of corrosion at a certain burnup, depending on the initial SPP size and fast neutron fluence. Only Zr alloys with irradiation resistant SPP avoid this type of increased corrosion completely. Two types of irradiation resistant materials were considered. One is a Zr-Sn-Fe alloy containing the Zr 3 Fe phase, which is irradiation resistant under BWR conditions. The other material is a Zr-Sn-Nb alloy containing the irradiation resistant β-Nb phase. In-BWR tests have shown that a Sn content of >0.8% is mandatory to minimize the nodular corrosion. Two prototypes of irradiation resistant alloys, Zr1.3Sn0.25-0.3 Fe and Zr1Sn2-3Nb, were irradiated in a BWR for 1372 days to a fast fluence of 9 x 10 21 n/cm 2 (E > 1 MeV). These irradiation tests showed that Zr1.3Sn0.25-0.3 Fe has a little lower resistance against nodular corrosion than optimized LTP (Low Temperature Process) Zircaloy-2/4 and revealed that Zr1Sn2-3Nb is superior to LTP Zircaloy-2/4 with respect to nodular and shadow corrosion resistance. The BWR corrosion resistance of Zr1Sn2-3Nb depends on heat treatment. The lowest corrosion was observed with material fabricated completely in the α-range, but also material manufactured in the lower (α+β)-range exhibits low corrosion. Material fabricated in the upper (α+β)-range showed a somewhat higher corrosion, a corrosion behavior similar to LTP Zircaloy-2/4. As far as final annealing is concerned, a long time annealing at 540 deg C is superior to a standard recrystallization treatment (e.g., at 580 deg C), which still leads to a corrosion behavior that is better than stress relieved Zr1Sn2-3Nb. Zr1Sn2-3Nb is resistant to shadow corrosion, when fabricated

  10. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

  11. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Smith, S.K.; Lehnert, D.F.; Locke, R.K.

    1991-01-01

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  12. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  13. Condensate treatment in BWR circuits by filter demineralizer units using powdered ion exchange resin at medium and high temperature

    International Nuclear Information System (INIS)

    De Martino, R.

    1983-01-01

    Considering the radiation build-up in some BWR reactors, we make a correlation between this phenomenon and the condensate purification system applied and the point of its utilization into the circuits. The application temperature of such a plant seems to have a very important role on the equilibria of metals contained in the reactor water and on the oxide composition. The efficiency of the condensate polishing system and the corrosion control are the most interesting objectives to achieve and to maintain, to control and regulate the physical and chemical process in the feedwater and in the reactor water. Up to date the technology owns major knowledge and a consistent know-how on using chemical products in order to increase the condensate polishing system efficiency. It is also considered a typical parallel case of a conventional power station and a secondary system of BWR units. (author)

  14. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  15. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1991-01-01

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR [boiling water reactor] in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed

  16. Leak before break analysis for cracking at multiple weld locations in BWR recirculation piping

    International Nuclear Information System (INIS)

    Zahoor, A.; Gamble, R.

    1984-01-01

    Periodically over the past decade, intergranular stress corrosion cracking (IGSCC) has been found in austenitic stainless steel piping at Boiling Water Reactor facilities. The effect of IGSCC on piping integrity has been evaluated previously in various BWR Owners Group and NRC studies. In these studies, the analyses were performed assuming the presence of a crack at a single weld location in the pipe run. The purpose of this investigation was to compare the leak rate and potential for unstable crack extension associated with a throughwall crack for the following two conditions in a BWR recirculation system: (1) the recirculation piping contains part through cracks at multiple weld locations and a single throughwall crack, and (2) the piping contains only a throughwall crack at one weld location. Two type BWRs were evaluated; namely, the ring header and five individual loop designs. The results from the analyses indicate that the potential for unstable crack extension at large bending loads, and leak rate at normal operation are not affected by the presence of part through cracks at multiple weld locations. The differences in the respective calculated L/sub eff/ and leak rates for the single and multiply cracked conditions are less than 2%

  17. Operation status display and monitoring system for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Wakabayashi, Yasuo; Hayakawa, Hiroyasu; Kawamura, Atsuo; Kaneda, Mitsunori.

    1982-01-01

    Lately, the development of the system has been made for BWR plants, which monitors the operating status not only in normal operation but also in abnormal state and also for plant safety. Recently, the improvement of man-machine interface has been tried through the practical use of technique which displays data collectively on a CRT screen relating them mutually. As one of those results, the practical use of an electronic computer and color CRT display for No. 1 unit in the Fukushima No. 2 Nuclear Power Station (2F-1), Tokyo Electric Power Co., is described. Also, new centralized control panels containing such systems were used for the 1100 MWe BWR nuclear power plants now under construction, No. 3 unit of the Fukushima No. 2 Power Station and No. 1 unit of Kashiwazaki-Kariwa Nuclear Power Station (2F-3 and K-1, respectively). The display and monitoring system in 2F-1 plant is the first one in which a computer and color CRTs were practically employed for a BWR plant in Japan, and already in commercial operation. The advanced operating status monitoring system, to which the result of evaluation of the above system was added, was incorporated in the new centralized control panels presently under production for 2F-3 and K-1 plants. The outline of the system, the functions of an electronic computer, plant operating status monitor, surveillance test guide, the automation of plant operation and auxiliary operation guide are reported for these advanced monitoring system. It was confirmed that these systems are useful means to improve the man-machine communication for plant operation minitoring. (Wakatsuki, Y.)

  18. Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

    International Nuclear Information System (INIS)

    Lucatero, Marco A.; Palacios-Hernandez, Javier C.; Ortiz-Villafuerte, Javier; Xolocostli-Munguia, J. Vicente; Gomez-Torres, Armando M.

    2010-01-01

    Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are

  19. The BWR VIP role in license renewal

    International Nuclear Information System (INIS)

    Dyle, R.

    2001-01-01

    The full text follows. The Boiling Water Reactor Vessel and Internals Project (BWRVIP) was started in response to an increase in the occurrence of intergranular stress corrosion cracking (IGSCC) in BWR reactor pressure vessel (RPV) internal components. The BWRVIP first evaluated the internals to determine which components were necessary to assure safe operation. An assessment of the relative significance of the internals was then performed to establish the priority in which components would be evaluated. Once this was determined, each safety-related component was evaluated to determine what, if any, inspections or tests were necessary to assure component integrity. Although IGSCC was the initial degradation mechanism of concern for RPV internals, the individual component evaluations considered all known modes of failure (fatigue, stress corrosion cracking, neutron embrittlement, etc). The component evaluations also considered all potential failure locations and the susceptibility to degradation. Once the potential failure locations and mechanisms were identified, the BWRVIP developed inspection criteria to assess component condition. The BWRVIP also developed flaw evaluation methodologies that could be used to determine the integrity and remaining life of each component. All of this information was consolidated into a document called an Inspection and Flaw Evaluation (IE) Guideline for each component. At the same time the BWRVIP was developing its program to assure internals integrity, utilities began to seriously consider measures necessary to extend the life of the plants. In the United States, the USNRC promulgated rules to allow the renewal of a license to allow plant operation for an additional 20 years. One aspect of the rule was that management of age-related degradation in the renewal period must be performed. The timing of this ''license renewal rule'' was advantageous in that it allowed the BWRVIP to address the requirements of the rule in the development

  20. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

  1. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  2. Limerick BWR turbine control and protection system upgrade success

    International Nuclear Information System (INIS)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A.; Williams, J.C.

    2015-01-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  3. Limerick BWR turbine control and protection system upgrade success

    Energy Technology Data Exchange (ETDEWEB)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A., E-mail: tangck@westinghouse.com, E-mail: pietryt@westinghouse, E-mail: federipa@westinghouse.com [Westinghouse Electric Company, LLC, Cranberry Township, PA (United States); Williams, J.C., E-mail: Jonathan.Williams@exeloncorp.com [Exelon Nuclear, Warrenville, IL (United States)

    2015-07-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  4. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  5. Modern technology applied in the advanced BWR (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.

    1988-01-01

    The advanced boiling water reactor (ABWR) represents the next generation of light water reactors (LWR) to be introduced into commercial operation in the 1990's. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. This paper discusses how the ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. A technical evaluation of the ABWR shows its superiority in terms of performance characteristics and economics relative to current LWR designs

  6. Evaluation of stress corrosion crack growth in BWR piping systems

    International Nuclear Information System (INIS)

    Kassir, M.; Sharma, S.; Reich, M.; Chang, M.T.

    1985-05-01

    This report presents the results of a study conducted to evaluate the effects of stress intensity factor and environment on the growth behavior of intergranular stress corrosion cracks in type 304 stainless steel piping systems. Most of the detected cracks are known to be circumferential in shape, and initially started at the inside surface in the heat affected zone near girth welds. These cracks grow both radially in-depth and circumferentially in length and, in extreme cases, may cause leakage in the installation. The propagation of the crack is essentially due to the influence of the following simultaneous factors: (1) the action of applied and residual stress; (2) sensitization of the base metal in the heat affected zone adjacent to girth weld; and (3) the continuous exposure of the material to an aggressive environment of high temperature water containing dissolved oxygen and some levels of impurities. Each of these factors and their effects on the piping systems is discussed in detail in the report. The report also evaluates the time required for hypothetical cracks in BWR pipes to propagate to their critical size. The pertinent times are computed and displayed graphically. Finally, parametric study is performed in order to assess the relative influence and sensitivity of the various input parameters (residual stress, crack growth law, diameter of pipe, initial size of defect, etc.) which have bearing on the growth behavior of the intergranular stress corrosion cracks in type 304 stainless steel. Cracks in large-diameter as well as in small-diameter pipes are considered and analyzed. 27 refs., 25 figs., 10 tabs

  7. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J M; Blazquez Martinez, J B

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  8. Filtered containment venting in Sweden

    International Nuclear Information System (INIS)

    Lindau, L.; Ellisson, K.

    1989-01-01

    After the TMI accident, Swedish authorities decided that all Swedish nuclear power plants should be upgraded with respect to mitigation of the consequences of severe accidents. One contribution to meet these ends is filtered containment venting, i.e. means to relieve containment overpressure and to clean the relief gas from radioactive components. The first system in operation was built at the Barsebaeck site (2 BWR's) where a gravel bed filter was installed. For the remaining Swedish units (7 BWR's and 3 PWR's) a passive, self-controlling wet scrubber system, FILTRA-MVSS, is now under installation. The principle of the FILTRA-MVSS is a self-controlling, self-pumping venturi collector submerged in a pool, and it is used to filter out emissions of solid and gaseous radioactive components

  9. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    reactor coolant system, combustible gas burning, atmosphere- structure heat transfer, ventilation, containment spray cooling, etc. After the molten core penetrates the reactor bottom head, steam generation, concrete disintegration and noncondensable gas generation are calculated in the reactor cavity or the pedestal. 2 - Method of solution: Each of the THALES member codes first establishes the steady state conditions after reading input data. Then iterative time-dependent calculation is continued, taking account of various phenomena and events and their interactions which will occur in the course of a postulated severe accident. The transient calculations are iterated by the physical times specified by input. Generally the RCS thermal hydraulic analysis with the THALES-PM or THALES-BM code is first carried out and its results are transferred to the following containment analysis with the THALES-CV code. Then both results are transferred to a code for analyzing fission product release and transport behavior. Automatic data transfer is possible in the case the JAERI's ART code is used for fission product behavior analysis. In overall thermal hydraulic analysis, a new method is adopted aiming at sufficiently accurate estimation of mixture levels in the reactor coolant system and the containment in a reasonable computer time. The heat transfer calculation in the core is carried out based on the backward method. 3 - Restrictions on the complexity of the problem: Restrictions relating to storage allocation are: (1) Maximum number of radial regions in the core : 10; (2) Maximum number of axial increments in the fuel rods : 50; (3) Maximum number of loops in the PWR primary system : 4; (4) Maximum number of volumes in the PWR primary system : 11; (5) Number of BWR recirculation loops: 2 (fixed); (6) Number of volumes in the BWR reactor coolant system : 7 (fixed); (7) Maximum number of compartments in the containment : 10. There is another restriction, which relates to time step

  10. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

  11. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  12. Requests on domestic nuclear data library from BWR design

    International Nuclear Information System (INIS)

    Maruyama, Hiromi

    2003-01-01

    Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

  13. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  14. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  15. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  16. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    Susilo, Jati

    2002-01-01

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  17. Physical model of nonlinear noise with application to BWR stability

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.

    1983-01-01

    Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD

  18. Power plant design: ESBWR - the latest passive BWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

    1997-01-01

    When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

  19. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  20. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  1. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  2. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  3. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    International Nuclear Information System (INIS)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment

  4. Assessment of the Prony's method for BWR stability analysis

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

    2011-01-01

    Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

  5. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  6. Laguna Verde nuclear power plant: an experience to consider in advanced BWR design

    International Nuclear Information System (INIS)

    Fuentes Marquez, L.

    2001-01-01

    Laguna Verde is a BWR 5 containment Mark II. Designed by GE, two external re-circulation loops, each of them having two speed re-circulation pump and a flow control valve to define the drive flow and consequently the total core flow an power control by total core flow. Laguna Verde Design and operational experience has shown some insights to be considering in design for advanced BRW reactors in order to improve the potential of nuclear power plants. NSSS and Balance of plant design, codes used to perform nuclear core design, margins derived from engineering judgment, at the time Laguna Verde designed and constructed had conducted to have a plant with an operational license, generating with a very good performance and availability. Nevertheless, some design characteristics and operational experience have shown that potential improvements or areas of opportunity shall be focused in the advanced BWR design. Computer codes used to design the nuclear core have been evolved relatively fast. The computers are faster and powerful than those used during the design process, also instrumentation and control are becoming part of this amazing technical evolution in the industry. The Laguna Verde experience is the subject to share in this paper. (author)

  7. Investigation of 3H and 14C inventory and distribution in spent BWR fuel rods

    International Nuclear Information System (INIS)

    Bleier, A.; Beuerle, M.; Neeb, K.H.

    1984-10-01

    In order to obtain reliable data for fuel reprocessing and waste disposal, the T and C-14 inventory, distribution and behaviour was investigated on a typical LWR fuel rod discharged from a BWR plant. The results showed that 50 ± 5% of the T generated in the fuel is present in the cladding after reactor operation. The remainder of the T stays with the fuel. Related to the reactor power the total T inventory corresponds to a T production rate of 19 000 Ci/GW e . a. The C-14 built up in the fuel represents approximately 60% of the C-14 inventory of the BWR fuel rod. The remaining part of C-14 (about 40%) experimentally determined by this analysis for the first time is generated in the cladding. From the total C-14 inventory a C-14 production rate of 17,5 Ci/GW e . a can be calculated. The fill gas contains only negligible fractions of both nuclides. The results obtained in this program are generally in good agreement with the data of theoretical estimates and with results of earlier investigations on PWR fuel rods. (orig.) [de

  8. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  9. Application of passive auto catalytic recombiner (PAR) for BWR plants in Japan

    International Nuclear Information System (INIS)

    Kobayashi, K.; Murano, K.; Yamanari, S.; Yamamoto, Y.

    2001-01-01

    The passive auto-catalytic recombiner (PAR), which can recombine flammable gases such as hydrogen and oxygen with each other to avoid an explosion in case of a loss-of-coolant accident (LOCA), installed in the primary containment vessel does not require a power supply or dynamic equipment, while the existing flammability gas control system (FCS) of most BWRs as an outer loop of the primary containment vessel needs them to make flammable gases circulate through blowers and heaters in the system. PAR offers a number of advantages over existing FCS, such as high reliability, low cost due to much smaller amount of materials needed, good maintainability, good operability in case of a LOCA, and smaller space for installation. An experimental study has been carried out for the purpose of solving the problems of applying PAR to Japanese BWR plants instead of existing FCS, in which we grasped the basic characteristics of PAR. (author)

  10. Effect of nitrogen in austenitic stainless steel on deformation behavior and stress corrosion cracking susceptibility in BWR simulated environment

    International Nuclear Information System (INIS)

    Roychowdhury, S.; Kain, V.; Dey, G.K.

    2012-01-01

    Intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) components in boiling water reactor (BWR has been a serious issue and is generic in nature. Initial cracking incidences were attributed to weld induced sensitisation and low temperature sensitisation which was mitigated by the use of low carbon grade of SS and molybdenum and nitrogen containing nuclear grade SS. However, IGSCC has occurred in these SS in the non-sensitised condition which was attributed to residual weld induced strain. Strain hardening in SS has been identified as a major cause for enhanced IGSCC susceptibility in BWR environment. Nitrogen in SS has a significant effect on the strain hardening characteristics and has potential to affect the IGSCC susceptibility in BWR environment. Type 304LN stainless steel is a candidate material for use in future reactors with long design life like the Advanced Heavy Water Reactor (AHWR), in which the operating conditions are similar to BWR. This study reports the effect of nitrogen in type 304LN stainless steel on the strain hardening behaviour and deformation characteristics and its effect on the IGSCC susceptibility in BWR/AHWR environment. Two heats of type 304LN stainless steel were used containing different levels of nitrogen, 0.08 and 0.16 wt % (SS alloys A and B, respectively). Both the SS was strain hardened by cross rolling at 200℃ to simulate the strain hardened regions having higher IGSCC susceptibility in BWRs. Tensile testing was done at both room temperature and 288℃(temperature simulating operating BWR conditions) and the effect of nitrogen on the tensile properties were established. Tensile testing was done at strain rates similar to the crack tip strain rates associated with a growing IGSCC in SS. Detailed transmission electron microscopic (TEM) studies were done to establish the effect of nitrogen on the deformation modes. Results indicated twinning was the major mode of deformation during cross rolling while

  11. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

  12. Impact of advanced BWR core physics method on BWR core monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

    2000-07-01

    Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

  13. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P.

    2006-01-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U 235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

  14. Mobile crud and transportation of radioactivity in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H-P. [Studsvik Nuclear AB, Nykoping (Sweden); LTU, Div. of Chemical Engineering, Lulea (Sweden); Hagg, J. [Ringhals AB, Varobacka (Sweden)

    2010-07-01

    Mobile crud is here referred to as a generic term for all types of particles that occur in the reactor water in BWRs and that are able to carry radioactivity. Previous results in this on-going series of studies in Swedish BWRs suggest that there are particles of different origins and function. A share may come from fuel crud and others may come from detachment, precipitation and dissolution processes in different parts of the BWR primary system, as well as from other system parts, such as the turbine/condenser. In addition, crud particles in this sense may come from purely mechanical processes such as degradation of graphite containing parts of the control rod drives. Therefore, the overall aim was to evaluate which particles are responsible for the transportation and distribution of radioactivity and also to clarify the chemical conditions under which they are formed. Furthermore the aim was to draw conclusions about how the chemistry would be like in order to avoid or at least minimize the formation of radioactivity distributing particles. A specific objective has also been to look into the importance of particle size for spreading of radioactivity in the primary system. Different types of crud particles are likely to have different characteristics in terms of function associated with transportation of radioactivity. The fuel crud is radioactive from the source and other types of crud can via surface processes, co-precipitation and other chemical and mechanical processes potentially affect the distribution of radioactivity in the primary system. In order to predict how operating parameters (e.g. stable, full power operation and scram) and chemical parameters (NWC/HWC/Zn, etc.) will affect the activity build-up on the system surfaces, it is important to know how the different types of crud are affected by these and related parameters. Fuel crud fixed on cladding ring samples, as well as mobile crud from the reactor water captured on filters, were examined by

  15. Comparative analysis of mechanical characteristics of solidified concentrates from BWR system using Yugoslav and Italian cements

    International Nuclear Information System (INIS)

    Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

    1987-01-01

    In this paper, properties of Italian and Yugoslav cement mixture with BWR evaporation concentrates were compared, research was held upon fifteen samples, according to the adequate formulations. Samples were made in standard cube form, side 10 cm. Functional relationship between decreasing the compressive strength and amount of incorporated BWR concentrate cement mixture was developed. The results of research showed nearly the same mechanical properties of solidified BWR concentrate with Italian and Yugoslav cements. (author)

  16. Kinematics of two-phase mixture level motion in BWR pressure vessels

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

    1985-01-01

    A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

  17. Construction techniques and management methods for BWR plants

    International Nuclear Information System (INIS)

    Shimizu, Yohji; Tateishi, Mizuo; Hayashi, Yoshishige

    1989-01-01

    Toshiba is constantly striving for safer and more efficient plant construction to realize high-quality BWR plants within a short construction period. To achieve these aims, Toshiba has developed and improved a large number of construction techniques and construction management methods. In the area of installation, various techniques have been applied such as the modularization of piping and equipment, shop installation of reactor internals, etc. Further, installation management has been upgraded by the use of pre-installation review programs, the development of installation control systems, etc. For commissioning, improvements in commissioning management have been achieved through the use of computer systems, and testing methods have also been upgraded by the development of computer systems for the recording and analysis of test data and the automatic adjustment of controllers in the main control system of the BWR. This paper outlines these construction techniques and management methods. (author)

  18. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  19. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    International Nuclear Information System (INIS)

    Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

    2008-01-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  20. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

    2008-07-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  1. A BWR 24-month cycle analysis using multicycle techniques

    International Nuclear Information System (INIS)

    Hartley, K.D.

    1993-01-01

    Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

  2. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

    Fukunishi, Kohyu

    1976-01-01

    The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

  3. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  4. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  5. Review of international solutions to NEACRP benchmark BWR lattice cell problems

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-12-01

    This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

  6. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  7. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  8. Results of modeling advanced BWR fuel designs using CASMO-4

    International Nuclear Information System (INIS)

    Knott, D.; Edenius, M.

    1996-01-01

    Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

  9. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

    2006-01-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

  10. TVA experience in BWR reload design and licensing

    International Nuclear Information System (INIS)

    Robertson, J.D.

    1986-01-01

    TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

  11. Non-local two phase flow momentum transport in S BWR

    International Nuclear Information System (INIS)

    Espinosa P, G.; Salinas M, L.; Vazquez R, A.

    2015-09-01

    The non-local momentum transport equations derived in this work contain new terms related with non-local transport effects due to accumulation, convection, diffusion and transport properties for two-phase flow. For instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where the local volume averaging equations fail. The S BWR was considered to study the non-local effects on the two-phase flow thermal-hydraulic core performance in steady-state, and the results were compared with the classical local averaging volume conservation equations. (Author)

  12. Non-local two phase flow momentum transport in S BWR

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Salinas M, L.; Vazquez R, A., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Apdo. Postal 55-535, 09340 Ciudad de Mexico (Mexico)

    2015-09-15

    The non-local momentum transport equations derived in this work contain new terms related with non-local transport effects due to accumulation, convection, diffusion and transport properties for two-phase flow. For instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where the local volume averaging equations fail. The S BWR was considered to study the non-local effects on the two-phase flow thermal-hydraulic core performance in steady-state, and the results were compared with the classical local averaging volume conservation equations. (Author)

  13. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  14. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  15. Sensitivity of BWR shutdown margin tests to local reactivity anomalies

    International Nuclear Information System (INIS)

    Cokinos, D.M.; Carew, J.F.

    1987-01-01

    Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

  16. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

  17. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  18. Advanced chemistry management system to optimize BWR chemistry control

    International Nuclear Information System (INIS)

    Maeda, K.; Nagasawa, K.

    2002-01-01

    BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

  19. Large bundle BWR test CORA-18: Test results

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

    1998-04-01

    The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

  20. Simulation of Thermal Stratification in BWR Suppression Pools with One Dimensional Modeling Method

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang

    2014-01-01

    The suppression pool in a boiling water reactor (BWR) plant not only is the major heat sink within the containment system, but also provides the major emergency cooling water for the reactor core. In several accident scenarios, such as a loss-of-coolant accident and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; the pool temperature distribution also affects the NPSHa (available net positive suction head) and therefore the performance of the Emergency Core Cooling System and Reactor Core Isolation Cooling System pumps that draw cooling water back to the core. Current safety analysis codes use zero dimensional (0-D) lumped parameter models to calculate the energy and mass balance in the pool; therefore, they have large uncertainties in the prediction of scenarios in which stratification and mixing are important. While three-dimensional (3-D) computational fluid dynamics (CFD) methods can be used to analyze realistic 3-D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, resulting in a long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code (Berkeley mechanistic MIXing code in C++) has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by one-dimensional (1-D) transient partial differential equations and substructures (such as free or wall jets) are modeled with 1-D integral models. This allows very large reductions in computational effort compared to multi-dimensional CFD modeling. One heat-up experiment performed at the Finland POOLEX facility, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, is used for

  1. An efficient modeling method for thermal stratification simulation in a BWR suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; Hua Li; Walter Villanueva; Pavel Kudinov

    2012-09-01

    The suppression pool in a BWR plant not only is the major heat sink within the containment system, but also provides major emergency cooling water for the reactor core. In several accident scenarios, such as LOCA and extended station blackout, thermal stratification tends to form in the pool after the initial rapid venting stage. Accurately predicting the pool stratification phenomenon is important because it affects the peak containment pressure; and the pool temperature distribution also affects the NPSHa (Available Net Positive Suction Head) and therefore the performance of the pump which draws cooling water back to the core. Current safety analysis codes use 0-D lumped parameter methods to calculate the energy and mass balance in the pool and therefore have large uncertainty in prediction of scenarios in which stratification and mixing are important. While 3-D CFD methods can be used to analyze realistic 3D configurations, these methods normally require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, therefore long simulation time. For mixing in stably stratified large enclosures, the BMIX++ code has been developed to implement a highly efficient analysis method for stratification where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. The POOLEX experiments at Finland, which was designed to study phenomena relevant to Nordic design BWR suppression pool including thermal stratification and mixing, are used for validation. GOTHIC lumped parameter models are used to obtain boundary conditions for BMIX++ code and CFD simulations. Comparison between the BMIX++, GOTHIC, and CFD calculations against the POOLEX experimental data is discussed in detail.

  2. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  3. Assembly Discontinuity Factors for the Neutron Diffusion Equation discretized with the Finite Volume Method. Application to BWR

    International Nuclear Information System (INIS)

    Bernal, A.; Roman, J.E.; Miró, R.; Verdú, G.

    2016-01-01

    Highlights: • A method is proposed to solve the eigenvalue problem of the Neutron Diffusion Equation in BWR. • The Neutron Diffusion Equation is discretized with the Finite Volume Method. • The currents are calculated by using a polynomial expansion of the neutron flux. • The current continuity and boundary conditions are defined implicitly to reduce the size of the matrices. • Different structured and unstructured meshes were used to discretize the BWR. - Abstract: The neutron flux spatial distribution in Boiling Water Reactors (BWRs) can be calculated by means of the Neutron Diffusion Equation (NDE), which is a space- and time-dependent differential equation. In steady state conditions, the time derivative terms are zero and this equation is rewritten as an eigenvalue problem. In addition, the spatial partial derivatives terms are transformed into algebraic terms by discretizing the geometry and using numerical methods. As regards the geometrical discretization, BWRs are complex systems containing different components of different geometries and materials, but they are usually modelled as parallelepiped nodes each one containing only one homogenized material to simplify the solution of the NDE. There are several techniques to correct the homogenization in the node, but the most commonly used in BWRs is that based on Assembly Discontinuity Factors (ADFs). As regards numerical methods, the Finite Volume Method (FVM) is feasible and suitable to be applied to the NDE. In this paper, a FVM based on a polynomial expansion method has been used to obtain the matrices of the eigenvalue problem, assuring the accomplishment of the ADFs for a BWR. This eigenvalue problem has been solved by means of the SLEPc library.

  4. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  5. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  6. Radiation streaming in power reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

    1979-02-01

    Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

  7. The use of the partial coherence function technique for the investigation of BWR noise dynamics

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

    The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

  8. K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1998-08-01

    This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k inf values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon consultation with the YMP staff, a Quad Cities BWR fuel assembly was selected as a baseline assembly. This design consists of seven axial enrichment zones, three of which contain natural uranium oxide. No attempt was made to find a bounding or even typical assembly design due to the wide variety in fuel assembly designs necessary for consideration. The current work concentrates on establishing a baseline analysis, along with a small number of sensitivity studies which can be expected later if desired. As a result of similar studies of this nature, several effects are known to be important in the determination of the final k inf for spent fuel in a cask-like geometry. For a given enrichment there is an optimal burnup: for lower burnups, excess energy (and corresponding excess reactivity) is present in the fuel assembly; for larger burnups, the assembly is overburned and essentially driven by neighboring fuel assemblies. The majority of the burnup/enrichment scenarios included in this study were for some near-optimum burnup/enrichment combinations as determined from Energy Information Administration (EIA) data. Several calculations were performed for under- and over-burned fuel to show these effects

  9. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

    2016-01-01

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

  10. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-08-15

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

  11. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    International Nuclear Information System (INIS)

    Gonzalez C, J.; Martin del Campo M, C.

    2003-01-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  12. BWR simulation in a stationary state for the evaluation of fuel cell design

    International Nuclear Information System (INIS)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A.

    2014-10-01

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  13. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  14. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  15. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  16. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

    2006-01-01

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

  17. Standard technical specifications General Electric plants, BWR/6. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/6 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  18. Standard technical specifications: General Electric plants, BWR/4. Volume 1, Revision 1: Specifications

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for General Electric BWR/4 Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.10 of the improved STS

  19. Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

    International Nuclear Information System (INIS)

    Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

    2002-01-01

    Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

  20. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  1. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user's manual

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs

  2. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  3. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    Hsiang-Shou Cheng; Diamond, D.J.

    1978-01-01

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  4. Measurements and analysis of neutron and gamma noise in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van; Kleiss, E.B.J.

    1985-01-01

    Neutron and gamma sensitive collectrons (self-powered detectors) have been designed for incore noise measurements in BWRs. A so-called twin-type has been developed for measurements of two-phase flow characteristics and detailed axial velocity distributions. Construction aspects of the twin detectors are discussed. An analysis is presented of the response of both detector types to incore parametric fluctuations. This analysis is based on detector response functions which provide an insight into the 'field of view' of the two types. The results are supported by experimental verifications; it is shown that incore gamma detectors provide useful additional information about two-phase flow in a BWR. (author)

  5. RETRAN experience with BWR transients at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

    1981-01-01

    Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

  6. A study on the feasibility of minor actinides in BWR

    International Nuclear Information System (INIS)

    Abdul Waris; Budiono

    2008-01-01

    Preliminary study on the feasibility of actinides minor (MA) recycling without mixing them with plutonium in boiling water reactor (BWR) has been carried out. The results show that increasing of fissile MA content in mixed oxide fuel (MOX) and/or reducing void fraction can enlarge the effective multiplication factor at the beginning of cycle, but the reactor still can not obtain its criticality condition. Furthermore, dropping the void fraction results in higher reactivity swing and therefore plummeting the safety factor of the reactor. (author)

  7. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  8. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

  9. The mechanical structure of the SVEA BWR fuel

    International Nuclear Information System (INIS)

    Nylund, O.; Johansson, A.; Junkrans, S.

    1985-01-01

    The SVEA BWR fuel assembly design is characterized by a double-wall cruciform internal structure forming an internal water gap and dividing the assembly into 4 subbundles. The effect is a favourable distribution of fuel and moderator, a minimum amount of structural material in active core, a combination of structural stability and flexibility for minimum control rod friction in reduced gaps and a reduced creep deformation of the fuel assembly. The results of a laboratory test program confirm the much lower friction force obtained with the SVEA fuel assemblies while withdrawing and inserting the control rod. (RF)

  10. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

    deRaad, J.A.; Wolters, J.T.

    1988-01-01

    Mechanized ultrasonic inspection of closure head studs often is on the critical path. In German BWR's, a floodcompensator is used which allows human access to the studs despite the water is up to a much higher level. For stud inspection this provides a potential solution to get out of the critical path. However, the space restrictions around the studs due to the geometry of the floodcompensator did not allow the use of the existing manipulators. This paper describes the design of a dedicated compact manipulator of a construction which copes with the restricted space available around the studs

  11. Decontamination and materials corrosion concerns in the BWR

    International Nuclear Information System (INIS)

    Gordon, B.M.; Gordon, G.M.

    1988-01-01

    The qualification of chemical decontamination processes to decontaminate complete systems or individual components in essential if effective inspection, maintenance, repair or replacement of plant components is to be achieved with minimum exposure of workers to ionizing radiation. However, it is critical that the benefits of decontamination processes are not overshadowed by deleterious materials/ corrosion side effects during the application of the process or during subsequent operation. This paper discusses such potential corrosion/materials problems in the BWR and presents relevant available corrosion data for the various commercial decontamination processes. (author)

  12. Characteristics of axial splits in failed BWR fuel rods

    International Nuclear Information System (INIS)

    Lysell, G.; Grigoriev, V.

    2000-01-01

    Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

  13. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  14. The noise analysis and the BWR operation map

    International Nuclear Information System (INIS)

    Blazquez, J.; Ballestrin, J.

    1996-01-01

    An analytical expression for the Decay Ratio is obtained: DR = exp(-bW / P 1/2 ). The physics behind is also explained. It applies to a commercial BWR Operation Map, on the vicinity of the power instability. This functional form seems fitting to the structure of the Operation map. The power P and the coolant flow are measured straightforward; the Decay Ratio is obtained by neutron noise analysis techniques. The parameter b, depending on the void reactivity coefficient, is then calculated on line during the Reactor Operation. New DR value is now predicted for each new displacement on the Map, so unexpected instability events are more likely avoided. (authors)

  15. Prediction of the stability of BWR reactors during the start-up process; Prediccion de la estabilidad de reactores BWR durante el proceso de arranque

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz E, J.A.; Castillo D, R. [ININ, Km. 36.5 Carretera Mexico-Toluca, 52045 Salazar, Estado de Mexico (Mexico); Blazquez M, J.B. [Centro de Investigaciones Energetics, Medioambientales y Tecnologicas, Av Complutense 22, 28040 Madrid (Spain)

    2004-07-01

    The Boiling Water Reactors (BWR) are susceptible of uncertainties of power when they are operated to low flows of coolant (W) and high powers (P), being presented this situation mainly in the start-up process. The start-up process could be made but sure if the operator knew the value of the stability index Decay reason (Dr) before going up power and therefore to guarantee the stability. The power and the flow are constantly measures, the index Dr could also be considered its value in real time. The index Dr depends on the power, flow and many other values, such as, the distribution of the flow axial and radial neutronic, the temperature of the feeding water, the fraction of holes and other thermohydraulic and nuclear parameters. A simple relationship of Dr is derived leaving of the pattern reduced of March-Leuba, where three independent variables are had that are the power, the flow and a parameter that it contains the rest of the phenomenology, that is to say all the other quantities that affect the value of Dr. This relationship developed work presently and verified its prediction with data of start-up of commercial reactors could be used for the design of a practical procedure practice of start-up, what would support to the operator to prevent this type of events of uncertainty. (Author)

  16. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  17. Core arrangement in BWR type reactors

    International Nuclear Information System (INIS)

    Asano, Masayuki.

    1981-01-01

    Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

  18. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  19. Big rock point restoration project BWR major component removal, packaging and shipping - planning and experience

    International Nuclear Information System (INIS)

    Milner, T.; Dam, S.; Papp, M.; Slade, J.; Slimp, B.; Nurden, P.

    2001-01-01

    The Big Rock Point boiling water reactor (BWR) at Charlevoix, MI was permanently shut down on August 29th 1997. In 1999 BNFL Inc.'s Reactor Decommissioning Group (RDG) was awarded a contract by Consumers Energy (CECo) for the Big Rock Point (BRP) Major Component Removal (MCR) project. BNFL Inc. RDG has teamed with MOTA, Sargent and Lundy and MDM Services to plan and execute MCR in support of the facility restoration project. The facility restoration project will be completed by 2005. Key to the success of the project has been the integration of best available demonstrated technology into a robust and responsive project management approach, which places emphasis on safety and quality assurance in achieving project milestones linked to time and cost. To support decommissioning of the BRP MCR activities, a reactor vessel (RV) shipping container is required. Discussed in this paper is the design and fabrication of a 10 CFR Part 71 Type B container necessary to ship the BRP RV. The container to be used for transportation of the RV to the burial site was designed as an Exclusive Use Type B package for shipment and burial at the Barnwell, South Carolina (SC) disposal facility. (author)

  20. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  1. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  2. Examination of minor actinide annihilation by BWR core

    International Nuclear Information System (INIS)

    Hida, Kazuki

    1995-01-01

    From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

  3. Appraisal of BWR plutonium burners for energy centers

    International Nuclear Information System (INIS)

    Williamson, H.E.

    1976-01-01

    The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

  4. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    International Nuclear Information System (INIS)

    Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

    2014-01-01

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  5. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  6. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.; Farmer, W.S.

    1992-01-01

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  7. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

  8. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)

  9. BWR Servicing and Refueling Improvement Program: Phase I summary report

    Energy Technology Data Exchange (ETDEWEB)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development.

  10. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1992-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  11. A study of heat capacity temperature limit of BWR

    International Nuclear Information System (INIS)

    Wang, Shih-Jen; Chen, Jyh-Jun; Chien, Chun-Sheng; Teng, Jyh-Tong

    2012-01-01

    Highlights: ► The purpose of this study is to verify the HCTL. ► MAAP4 was used as code to generate a realistic and convenient HCTL. ► The current HCTL curve causes confusing in reading data. ► The revised HCTL curves developed in this study. ► Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners’ group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  12. System control model of a turbine for a BWR

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A.

    2009-10-01

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  13. Transmutation of minor actinide using BWR fueled mixed oxide

    International Nuclear Information System (INIS)

    Susilo, Jati

    2000-01-01

    Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

  14. BWR Servicing and Refueling Improvement Program: Phase I summary report

    International Nuclear Information System (INIS)

    Perry, D.R.

    1978-09-01

    Under the U.S. Department of Energy sponsorship, General Electric Co. (GE) undertook a study of boiling water reactor (BWR) refueling outages for the purpose of recommending the development and demonstration of critical path time savings improvements. The Tennessee Valley Authority (TVA) joined the study as a subcontractor, providing monitoring assistance and making the Browns Ferry Site available for improvement demonstrations. Agreement was also reached with Georgia Power Co., Power Authority of the State of New York, and Commonwealth Edison Co. for monitoring and data collection at Hatch 1, FitzPatrick, and Quad Cities 1 nuclear plants, respectively. The objective was to identify, develop, and demonstrate improved refueling, maintenance, and inspection procedures and equipment. The improvements recommended in this study are applicable to BWR nuclear plants currently in operation as well as those in the design and construction phases. The recommendations and outage information can be used as a basis to plan and conduct the first outages of new plants and to improve the planning and facilities of currently operating plants. Many of the recommendations can readily be incorporated in plants currently in the design and construction phases as well as in the design of future plants. Many of these recommended improvements can be implemented immediately by utilities without further technical development

  15. Evaluation of thermal margin during BWR neutron flux oscillation

    International Nuclear Information System (INIS)

    Takeuchi, Yutaka; Takigawa, Yukio; Chuman, Kazuto; Ebata, Shigeo

    1992-01-01

    Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three-dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio). The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit. A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide -oscillation, because of the sensitivity difference in a flow-to-power gain. A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals. (author)

  16. Safety evaluation of BWR off-gas treatment systems

    International Nuclear Information System (INIS)

    Schultz, R.J.; Schmitt, R.C.

    1975-01-01

    Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)

  17. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This Phase 1 Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assesses the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of NPRDS failure cases attributed to the CRD system, and (4) personal information exchange. As part of this study, nearly 3,500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation has been conducted that summarizes the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented regarding specific actions that utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities

  18. A study of heat capacity temperature limit of BWR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shih-Jen, E-mail: sjenwang@iner.gov.tw [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Chen, Jyh-Jun [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China); Chien, Chun-Sheng [Institute of Nuclear Energy Research (INER), 1000, Wunhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd., Chung Li City, Taoyuan County 32023, Taiwan (China)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer The purpose of this study is to verify the HCTL. Black-Right-Pointing-Pointer MAAP4 was used as code to generate a realistic and convenient HCTL. Black-Right-Pointing-Pointer The current HCTL curve causes confusing in reading data. Black-Right-Pointing-Pointer The revised HCTL curves developed in this study. Black-Right-Pointing-Pointer Users can obtain important parameters from the revised HCTL without confusion and interpolation. - Abstract: Heat capacity temperature limit (HCTL) is an important parameter for operation of BWR. Current version of the HCTL was derived, based on simple model of computation aids (CA) of BWR owners' group (BWROG). However, some parts of the current HCTL are confusing to the users in reading data. The purpose of this study is to verify the HCTL by applying the MAAP4 code to the field of emergency operating procedure (EOP). The trends of HCTL generated by MAAP4 code are consistent with those obtained from CA. A series of revised HCTL evaluated at various times after scram are provided and the confusing part is eliminated.

  19. A detailed BWR recirculation loop model for RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

    2017-01-15

    Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

  20. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  1. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  2. Operational experiences with on line BWR condenser tube leak verification

    International Nuclear Information System (INIS)

    Bryant, R.A.; Duvall, W.E.; Kirkley, W.B.; Zavadoski, R.W.

    1988-01-01

    Verifying condenser tube leaks at a boiling water reactor is, at best, a difficult task carried out in hot steamy water boxes with concurrent radiation exposure. For small apparent leaks with slight chemical changes there is always uncertainty of whether the problem is a condenser tube leak or a feedback from radwaste. Most conventional methods (e.g soap tests, Saran wrap suction, and helium tests) usually involve a load reduction to isolate the water boxes one at a time and hours of drain down on each box. The sensitivity of the most sensitive test (helium) is of the order of 7500 l per day per box. Sulfur hexafluoride has been successfully used at a BWR to identify one leaking water box out of four while the unit was at 100 % power. The actual tubes leakig in the water box were identified by injecting helium during drain down of the box and subsequent manifold testing. Additional tests with sulfur hexafluoride on the second BWR unit indicated tight water boxes to within the sensitivity of the measurement, i.e. less than 19 l per day for all four boxes. Problems encountered in both tests included sulfur hexafluoride carry over from the plume of the cooling towers and off gas considerations. In brief sulfur hexafluoride can be used to quickly identify which particular water box has a condenser tube leak or, just as quickly, establish the integrity of all the water boxes to a level not previously attainable. (author)

  3. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

    Rajamaeki, Markku.

    1980-03-01

    TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

  4. Logical model for the control of a BWR turbine

    International Nuclear Information System (INIS)

    Vargas O, Y.; Amador G, R.; Ortiz V, J.; Castillo D, R.

    2009-01-01

    In this work a design of a logical model is presented for the turbine control of a nuclear power plant with a BWR like energy source. The model is sought to implement later on inside the thermal hydraulics code of better estimate RELAP/SCDAPSIM. The logical model is developed for the control and protection of the turbine, and the consequent protection to the BWR, considering that the turbine control will be been able to use for one or several turbines in series. The quality of the present design of the logical model of the turbine control is that it considers the most important parameters in the operation of a turbine, besides that they have incorporated to the logical model the secondary parameters that will be activated originally as true when the turbine model is substituted by a detailed model. The development of the logical model of a turbine will be of utility in the short and medium term to carry out analysis on the turbine operation with different operation conditions, of vapor extraction, specific steps of the turbine to feed other equipment s, in addition to analyze the separate and the integrated effect. (Author)

  5. Containment vessel design and practice

    International Nuclear Information System (INIS)

    Bangash, Y.

    1983-01-01

    The state of the art of analysis and design of the concrete containment vessels required for BWR and PWR is reviewed. A step-by-step critical appraisal of the existing work is given. Elastic, inelastic and cracking conditions under extreme loads are fully discussed. Problems associated with these structures are highlighted. A three-dimensional finite element analysis is included to cater for service, overload and dynamic cracking of such structures. Missile impact and seismic effects are included in this work. The second analysis is known as the limit state analysis, which is given to design such vessels for any kind of load. (U.K.)

  6. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  7. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  8. 3D modeling of missing pellet surface defects in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

    2016-10-15

    Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

  9. Dyre-kloning og marked i USA

    DEFF Research Database (Denmark)

    Meyer, Gitte

    2007-01-01

    Et dilemma presser sig på: hvis kun videnskabeligt dokumenterede risici kan begrunde mærkningskrav, mens alle andre etiske spørgsmål er private – hvordan skal amerikanske forbrugere så bære sig ad med at praktisere deres private etik?...

  10. Nuclear reactor container

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1989-01-01

    Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

  11. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.

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

    Science.gov (United States)

    Artnak, Edward Joseph, III

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

  13. System control model of a turbine for a BWR; Modelo del sistema de control de una turbina para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, Y. [Universidad del Valle de Mexico, Campus Toluca, Av. Las Palmas No. 136, Col. San Jorge Pueblo Nuevo, 52140 Metepec, Estado de Mexico (Mexico); Amador G, R.; Ortiz V, J.; Castillo D, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rodolfo.amador@inin.gob.mx

    2009-10-15

    In this work is presented a design of a control system of a turbine for a nuclear power plant with a BWR like energy source. The model seeks to implement later on at thermal hydraulics code of better estimate RELAP/SCDAPSIM. The model is developed for control and protection of turbine, and the consequent protection to the BWR, considering that the turbine control could be employed for one or several turbines in series. The quality of present designs of control pattern of turbine it is that it considers the parameters more important in the operation of a turbine besides that is has incorporated at control the secondary parameters that will be activated originally as true when the turbine model is substituted by a model more detailed. The development of control model of a turbine will be good in short and medium term to realize analysis about the operation of turbine with different operation conditions, of vapor extraction specific steps of turbine to feed other equipment s, besides analyzing the separate effect and integrated effect. (Author)

  14. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  15. FIST small break accident analysis with BWR TRACBO2-pretest predictions

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

  16. Water chemistry control practices and data of the European BWR fleet

    International Nuclear Information System (INIS)

    Stellwag, B.; Laendner, A.; Weiss, S.; Huettner, F.

    2010-01-01

    Nineteen BWR plants are in operation in Europe, nine built by ASEA Atom, six by Siemens KWU and four by General Electric. This paper gives an overview of water chemistry operation practices and parameters of the European BWR plants. General design characteristics of the plants are described. Chemistry control strategies and underlying water chemistry guidelines are summarized. Chemistry data are presented and discussed with regard to plant design characteristics. The paper is based on a contract of the European BWR Forum with AREVA on a chemistry sourcebook for member plants. The survey of chemistry data was conducted for the years 2002 to 2008. (author)

  17. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

  18. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1992-03-01

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  19. Automatic determination of BWR fuel loading patterns based on K.E. technique with core physics simulation

    International Nuclear Information System (INIS)

    Ikehara, T.; Tsuiki, M.; Takeshita, T.

    1990-01-01

    On the basis oof a computerized search method, a prototype for a fuel loading pattern expert system has been developed to support designers in core design for BWRs. The method was implemented by coupling rules and core physics simulators into an inference engine to establish an automated generate-and-test cycle. A search control mechanism, which prunes paths to be searched and selects appropriate rules through the interaction with the user, was also introduced to accomplish an effective search. The constraints in BWR core design are: (1) cycle length more than L, (2) core shutdown margin more than S, and (3) thermal margin more than T. Here L, S, and T are the specified minimum values. In this system, individual rules contain the manipulation to improve the core shutdown margin explicitly. Other items were taken into account only implicitly. Several applications to the test cases were carried out. It was found that the results were comparable with those obtained by human expert engineers. Broad applicability of the present method in the BWR core design domain was proved

  20. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor

    International Nuclear Information System (INIS)

    Victoria R, M.A.; Morales S, J.B.

    2005-01-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  1. Instrumentation and testing of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Pace, D.W.; Klamerus, E.W.

    1997-01-01

    Static overpressurization tests of two scale models of nuclear containment structures - a steel containment vessel (SCV) representative of an improved, boiling water reactor (BWR) Mark II design and a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR) - are being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. This paper discusses plans for instrumentation and testing of the PCCV model. 6 refs., 2 figs., 2 tabs

  2. Global vibrations in the wetwell condensation process caused by LOCA in BWR plants

    International Nuclear Information System (INIS)

    Bjoerndahl, O.; Andersson, Magnus

    1998-12-01

    During the last years a substantial part of third part review work related to dynamical loadings has been review of loading specifications dealing with vibrations in containment building related to so called LOCA-events in Swedish BWR plants. Compared to other loading categories characterised as global vibrations these secondary effects of LOCA-events are complex to analyse. One experience from the review work at SAQ up to now is that it is not fully clear what prediction methods and what model idealisations are the most adequate for structural integrity verification on mechanical systems as pressure vessels and piping under such loading conditions. At SAQ Teknik a project work has been carried out to investigate the general status of the methodology used today in Sweden and a work to in the long term develop simplified prediction models and methods for the loading categories condensation oscillations (CO) and chugging (CH). The work was initially concentrated on a study of the background of the methodology which was developed for these type of loading in American BWR-containments of the Mark-II design. The methodology was developed by General Electric, GE, in cooperation with the Mark-II plant owners. The methodology used in Sweden to predict vibrations in BWR containments of this design is with some minor modifications very close to technique developed by GE. The methodology developed by GE is the only accepted by USNRC for the Mark-II design and could be found as reference in Standard Review Plan 6.2.1.1.C, Rev 6 - August 1984. Based on identical physical assumptions about the dynamic behaviour of the building structure and the water in the suppression pool, mathematical models are derived in this report for predictions of secondary structure response spectra for loading conditions as global vibrations during CO and CH. Based on parameters identified by so called one pipe experiments responses my be predicted. By use of these derived mathematical models as a

  3. Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

    International Nuclear Information System (INIS)

    Park, Soo Young; Ahn, Kwang Il

    2012-01-01

    Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

  4. Modeling of SCC initiation and propagation mechanisms in BWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmeister, Hans, E-mail: Hans.Hoffmeister@hsu-hh.de [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany); Klein, Oliver [Institute for Failure Analysis and Failure Prevention ISSV e.V., c/o Helmut Schmidt University of the Federal Armed Forces, D-22039 Hamburg (Germany)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We show that SSC in BWR environments includes anodic crack propagation and hydrogen assisted cracking. Black-Right-Pointing-Pointer Hydrogen cracking is triggered by crack tip acidification following local impurity accumulations and subsequent phase precipitations. Black-Right-Pointing-Pointer We calculate effects of pH, chlorides, potentials and stress on crack SCC growth rates at 288 Degree-Sign C. - Abstract: During operation of mainly BWRs' (Boiling Water Reactors) excursions from recommended water chemistries may provide favorite conditions for stress corrosion cracking (SCC). Maximum levels for chloride and sulfate ion contents for avoiding local corrosion are therefore given in respective water specifications. In a previously published deterministic 288 Degree-Sign C - corrosion model for Nickel as a main alloying element of BWR components it was demonstrated that, as a theoretically worst case, bulk water chloride levels as low as 30 ppb provide local chloride ion accumulation, dissolution of passivating nickel oxide and precipitation of nickel chlorides followed by subsequent local acidification. In an extension of the above model to SCC the following work shows that, in a first step, local anodic path corrosion with subsequent oxide breakdown, chloride salt formation and acidification at 288 Degree-Sign C would establish local cathodic reduction of accumulated hydrogen ions inside the crack tip fluid. In a second step, local hydrogen reduction charges and increasing local crack tip strains from increasing crack lengths at given global stresses are time stepwise calculated and related to experimentally determined crack critical cathodic hydrogen charges and fracture strains taken from small scale SSRT tensile tests pieces. As a result, at local hydrogen equilibrium potentials higher than those of nickel in the crack tip solution, hydrogen ion reduction initiates hydrogen crack propagation that is enhanced with

  5. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network

    International Nuclear Information System (INIS)

    Montes, J.L.; Ortiz, J.J.; Perusquia C, R.; Francois, J.L.; Martin del Campo M, C.

    2007-01-01

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U 235 , some of these bars also contain a concentration of Gd 2 O 3 and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  6. Testing of a steel containment vessel model

    International Nuclear Information System (INIS)

    Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.; Komine, K.; Costello, J.F.

    1997-01-01

    A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

  7. Study of transient turbine shot without bypass in a BWR

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    The study and analysis of operational transients are important for predicting the behavior of a system to short-terms events and the impact that would cause this transition. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could result in an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis of results of a turbine shot transient, which is not taken into operation the bypass is presented. The study is realized for a BWR of 2027 MWt, to an intermediate cycle life and using the computer code Simulate-3K a depressurization stage of the vessel is created which shows the response of other security systems and gives a coherent prediction to the event presented type. (Author)

  8. Core concept for long operating cycle simplified BWR (LSBWR)

    International Nuclear Information System (INIS)

    Kouji, Hiraiwa; Noriyuki, Yoshida; Mikihide, Nakamaru; Hideaki, Heki; Masanori, Aritomi

    2002-01-01

    An innovative core concept for a long operating cycle simplified BWR (LSBWR) is currently being developed under a Toshiba Corporation and Tokyo Institute of Technology joint study. In this core concept, the combination of enriched uranium oxide fuels and loose-pitched lattice is adopted for an easy application of natural circulation. A combination of enriched gadolinium and 0.7-times sized small bundle with peripheral-positioned gadolinium rod is also adopted as a key design concept for 15-year cycle operation. Based on three-dimensional nuclear and thermal hydraulic calculation, a nuclear design for fuel bundle has been determined. Core performance has been evaluated based on this bundle design and shows that thermal performance and reactivity characteristics meet core design criteria. Additionally, a control rod operation plan for an extension of control rod life has been successfully determined. (author)

  9. Cobra-TF simulation of BWR bundle dry out experiments

    Energy Technology Data Exchange (ETDEWEB)

    Frepoli, C.; Ireland, A.; Hochreiter, L.; Ivanov, K. [Penn State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, PA (United States); Velten, R. [Siemens Nuclear Power GmbH, Erlangen (Germany)

    2001-07-01

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. This study presents the results obtained with COBRA-TF for the simulation of the Siemens 9-9Q BWR Bundle Dryout experiments. The model includes 20 channels and 34 axial nodes in the heated section. The predicted critical power and dryout location is compared with the measured values. An assessment of the code entrainment and de-entrainment models is presented. (authors)

  10. BWR Full Integral Simulation Test (FIST). Phase I test results

    International Nuclear Information System (INIS)

    Hwang, W.S.; Alamgir, M.; Sutherland, W.A.

    1984-09-01

    A new full height BWR system simulator has been built under the Full-Integral-Simulation-Test (FIST) program to investigate the system responses to various transients. The test program consists of two test phases. This report provides a summary, discussions, highlights and conclusions of the FIST Phase I tests. Eight matrix tests were conducted in the FIST Phase I. These tests have investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. Results and governing phenomena of each test have been evaluated and discussed in detail in this report. One of the FIST program objectives is to assess the TRAC code by comparisons with test data. Two pretest predictions made with TRACB02 are presented and compared with test data in this report

  11. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Watanabe, Norio

    1982-06-01

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10 - 3 /demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  12. Development of methodology for early detection of BWR instabilities

    International Nuclear Information System (INIS)

    Alessandro Petruzzi; Shin Chin; Kostadin Ivanov; Asok Ray; Fan-Bill Cheung

    2005-01-01

    Full text of publication follows: The objective of the work presented in this paper research, which is supported by the US Department of Energy under the NEER program, is to develop an early anomaly detection methodology in order to enhance safety, availability, and operational flexibility of Boiling Water Reactor (BWR) nuclear power plants. The technical approach relies on suppression of potential power oscillations in BWRs by detecting small anomalies at an early stage and taking appropriate prognostic actions based on an anticipated operation schedule. The model of coupled (two-phase) thermal-hydraulic and neutron flux dynamics, based on the US NRC coupled code TRACE/PARCS, is being utilized as a generator of time series data for anomaly detection at an early stage. The concept of the methodology is based on the fact that nonlinear systems show bifurcation, which is a change in the qualitative behavior as the system parameters vary. Some of these parameters may change on their own accord and account for the anomaly, while certain parameters can be altered in a controlled fashion. The non-linear, non-autonomous BWR system model considered in this research exhibits phenomena at two time scales. Anomalies occur at the slow time scale while the observation of the dynamical behavior, based on which inferences are made, takes place at the fast time scale. It is assumed that: (i) the system behavior is stationary at the fast time scale; and (ii) any observable non-stationary behavior is associated with parametric changes evolving at the slow time scale. The goal is to make inferences about evolving anomalies based on the asymptotic behavior derived from the computer simulation. However, only sufficient changes in the slowly varying parameter may lead to detectable difference in the asymptotic behavior. The need to detect such small changes in parameters and hence early detection of an anomaly motivate the utilized stimulus-response approach. In this approach, the model

  13. Investigation of detector tube impacting in the Ringhals-1 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, C.; Pazsit, I. [Department of Nuclear Engineering, Chalmers University of Technology, SE-412 96 Goteborg (Sweden)]. E-mail: kalle@nephy.chalmers.se; imre@nephy.chalmers.se

    2006-07-01

    Neutron noise measurements were made in two consecutive fuel cycles in the Swedish BWR Ringhals-1 with the purpose of diagnostics of vibrations and impacting of detector strings. Two diagnostic tools were used, first a traditional spectral analysis and second a wavelet-based method. Both types of methods have been used in the past, but not simultaneously during one fuel cycle. In addition, a new method, wavelet-based coherence, was tested with success. Based on the results of the analysis, with emphasis on the traditional method, the detector tubes were divided into three groups with respect to the severity and likelihood of impacting. For the first series of measurements, these conclusions could be checked against visual inspection of the fuel assemblies during refuelling after the cycle, in order to find impacting damage. A good correlation between the prediction of the analysis and the inspection results was found. (author)

  14. Investigation of detector tube impacting in the Ringhals-1 BWR

    International Nuclear Information System (INIS)

    Sunde, C.; Pazsit, I.

    2006-01-01

    Neutron noise measurements were made in two consecutive fuel cycles in the Swedish BWR Ringhals-1 with the purpose of diagnostics of vibrations and impacting of detector strings. Two diagnostic tools were used, first a traditional spectral analysis and second a wavelet-based method. Both types of methods have been used in the past, but not simultaneously during one fuel cycle. In addition, a new method, wavelet-based coherence, was tested with success. Based on the results of the analysis, with emphasis on the traditional method, the detector tubes were divided into three groups with respect to the severity and likelihood of impacting. For the first series of measurements, these conclusions could be checked against visual inspection of the fuel assemblies during refuelling after the cycle, in order to find impacting damage. A good correlation between the prediction of the analysis and the inspection results was found. (author)

  15. TRAB - A transient analysis program for BWR. Part 2

    International Nuclear Information System (INIS)

    Raety, H.; Rajamaeki, M.

    1991-05-01

    TRAB is a transient analysis code for BWRs developed at the Technical Research Centre of Finland. It models the phenomena in the interior of the BWR pressure vessel and in related subsystems. The core model of TRAB can be used separately for LWR modelling. For PWR modelling the core model of TRAB is connected to circuit model SMABRE to form the SMATRA code. This report is a user's manual and documents the structure, contents and preparation of input for TRAB. The structure of TRAB input is very flexible, featuring input groups and subgroups identified with keywords and given in any order as well as data items in free format, freely mixed with explanatory texts. Users interface of the code can be used for modelling within input: through normal input it is possible to create new submodels. These may be functional or tabulated dependencies of the code variables, different types of delays, or ordinary linear differential equations

  16. Seismic response analysis of BWR buildings with embedded foundation

    International Nuclear Information System (INIS)

    Fukuzawa, R.; Chiba, O.; Tohdo, M.

    1985-01-01

    An investigation on the effect of various types of foundations embedded into soil deposit to the responses of BWR buildings is made. The frequency characteristics of massless rigid foundations such as impedance functions and input motion are first obtained, depending on the embedment depth, contact condition of side wall to soil and the thickness of surface layer in a two-layered soil. From the results it is found that the embedment depth and the thickness of surface layer have a distinct effect to the responses contributed by the impedance functions and the response values characterized by the input motions are obviously influenced by the embedment depth and the contact length of side wall to soil deposit. (orig.)

  17. Siemens Nuclear Power Corporation experience with BWR and PWR fuels

    International Nuclear Information System (INIS)

    Reparaz, A.; Smith, M.H.; Stephens, L.G.

    1992-01-01

    The large data base of fuel performance parameters available to Siemens Nuclear Power Corporation (SNP), and the excellent track record of innovation and fuel reliability accumulated over the last twenty-three years, allows SNP to have a clear insight on the characteristics of future developments in the area of fuel design. Following is a description of some of SNP's recent design innovations to prevent failures and to extend burnup capabilities. A goal paramount to the design and manufacture of BWR and PWR fuel is that of zero defects from any case during its operation in the reactor. Progress has already been made in achieving this goal. This paper summarized the cumulative failure rate of SNP fuel rod through January 1992

  18. An ecological interface design for BWR nuclear power plants

    International Nuclear Information System (INIS)

    Monta, K.; Itoh, J.

    1992-01-01

    An ecological interface design was applied to realize the support function for the operator's direct perception and analytical reasoning in the development of an intelligent man-machine system for BWR nuclear power plants. The abstraction-aggregation functional hierarchy representation of the work domain is a base of the ecological interface design. Another base is the concept of the level of cognitive control. The former was mapped into the interface to externalize the operator's normative mental model of the plants, which will reduce his/her cognitive work load and support knowledge-based problem solving. In addition, the same framework can be used for the analytical evaluation of man-machine interfaces. The information content and structure of a prototype interface were evaluated. This approach seems promising from these experiences. (author)

  19. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  20. Availability analysis of United States BWR IV electrical generation plants

    International Nuclear Information System (INIS)

    Renick, D.H.; Li, F.; Todreas, N.E.

    1998-01-01

    Availability, as quantified by power output levels, from all active U.S. BWR IV plants were analyzed over a seven and a half year period to determine the operational characteristics of these plants throughout an operating cycle. The operational data were examined for infant mortality, end of cycle decreased availability, and seasonal availability variations. Scheduled outages were also examined to determine the industry's current approach to planning maintenance outages. The results of this study show that nuclear power plants do suffer significant infant mortality following a refueling outage. And while they do not suffer an end of cycle decrease in availability, a mid-cycle period of decreased availability is evident. This period of decreased availability is due to a combination of increased forced unavailability and seasonally scheduled maintenance and refueling outages. These findings form the start of a rational approach to increasing plant availability. (author)