Reference Pressurized Water Reactor Research Articles

Analyses of Critical Hydrogen Enrichment in PWR Containment Compartments

During a loss-of-coolant accident in a pressurized water reactor (PWR), steam of varying quality is released from the primary circuit into the equipment compartments of the containment, followed by the release of a hydrogen-steam mixture during the core degradation phase. In the case of long-lasting accidents, findings of detailed code analyses indicate an enrichment of hydrogen in lower peripheral containment compartments in the reference PWR plant under investigation. During the late accident phase with ex-vessel molten core–concrete interaction, even in the case of an operating passive autocatalytic recombiner system, this poses a threat for local hydrogen combustion later on. Such hydrogen phenomena are not expected and have not been widely studied up to now. Therefore, corresponding experiments have been performed at the THAI test facility operated by Becker Technologies. One of these tests had been precalculated with the COntainment COde SYStem (COCOSYS) as part of the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) code system AC2 and has been used to validate the code. The 60-m3 THAI test vessel has been divided into an inner compartment that has been connected to the surrounding vessel, simulating the upper and peripheral containment part, by very small flow openings at the bottom representing the clearance between door frames and door leaves and one opening at the top representing typical openings by burst disks. The paper discusses both the experimental findings of a test series on the potential enrichment of hydrogen in lower containment compartments and the COCOSYS calculations demonstrating the applicability of the code under complex flow conditions including stratification phenomena.

Effect of mitigation strategies in the severe accident uncertainty analysis of the OPR1000 short-term station blackout accident

Integrated severe accident codes should be capable of simulating not only specific physical phenomena but also entire plant behaviors, and in a sufficiently fast time. However, significant uncertainty may exist owing to the numerous parametric models and interactions among the various phenomena. The primary objectives of this study are to present best-practice uncertainty and sensitivity analysis results regarding the evolutions of severe accidents (SAs) and fission product source terms and to determine the effects of mitigation measures on them, as expected during a short-term station blackout (STSBO) of a reference pressurized water reactor (optimized power reactor (OPR)1000). Three reference scenarios related to the STSBO accident are considered: one base and two mitigation scenarios, and the impacts of dedicated severe accident mitigation (SAM) actions on the results of interest are analyzed (such as flammable gas generation). The uncertainties are quantified based on a random set of Monte Carlo samples per case scenario. The relative importance values of the uncertain input parameters to the results of interest are quantitatively evaluated through a relevant sensitivity/importance analysis.

Best-practice severe accident uncertainty and sensitivity analysis for a short-term SBO sequence of a reference PWR using MAAP5

The integrated severe accident codes like MELCOR and MAAP5 include numerous uncertain models and parameters representing a wide range of physico-chemical phenomena for a comprehensive plant simulation, consequently making relevant uncertainty analysis a non-trivial task. In this paper, best-practice uncertainty and sensitivity analysis was performed to statistically quantify uncertainties associated with the analysis results of interest and identify potentially important uncertain parameters, based on the key modeling parameters employed in MAAP5. Three reference scenarios related to a short-term station blackout (STSBO) accident of a reference pressurized water reactor (PWR: OPR1000), were considered for the purpose: one base scenario and two mitigation scenarios to investigate the impact of dedicated severe accident mitigation (SAM) actions on the results of interest. A series of uncertainty and sensitivity analyses were carried out for these model parameters. Uncertainties for the results of interest were quantified through a random set of the Monte Carlo samples per case scenario, with values statistically sampled from the probability distributions of the individual model parameters. In addition, the relative importance of these model parameters to each relevant analysis result was then quantitatively evaluated through the sensitivity and importance analysis. Analysis results and relevant insights are summarized in terms of particular points of interest.

Plant-specific assessment of the natural circulation-induced creep rupture characteristics in the RCS pressure boundary during the SBO accident

In nuclear power plants (NPPs), one of the important yet uncertain phenomena during a station blackout (SBO) accident is the possibility for a counter-current natural circulation flow in the reactor coolant system (RCS) pressure boundary, which could divert the core melt progression and/or the in-vessel fission product (FP) distribution and, in turn, lead to a heatup and eventual creep rupture. From the risk and accident management points of view, one of key concerns is to predict more explicitly whether such a creep rupture can occur in the RCS pressure boundary, and if it indeed can, also predict the most probable time and the most likely location. This paper presents plant-specific analysis results on the potential characteristics of natural circulation-induced creep rupture of the RCS pressure boundary that could potentially take place during an SBO accident of a reference pressurized water reactor (APR1400). MELCOR-CFD (Computational Fluid Dynamics) coupled analysis was performed to determine more precisely the relevant physical parameter that could greatly influence on the behaviour of natural circulation and the ensuing creep rupture. Multiple sensitivity analyses were also carried out to investigate the influence that particular MELCOR modeling schemes and plant conditions have in estimating the RCS creep rupture characteristics. The analysis results and relevant insights were summarized in terms of particular points of interest.

Best-practice severe accident analysis for the OPR1000 short-term SBO sequence using MELCOR2.2 and MAAP5

A short-term station blackout (STSBO) accident has been considered as one of risk dominant sequences for nuclear power plants (NPPs). Unless appropriate mitigation actions are not taken, a rapid damage of the reactor core and thereby a severe accident may result in. This paper presents best practice analysis results on the evolution of the severe accident sequences and fission product (FP) source terms expected during the STSBO accident of a reference pressurized water reactor. In addition, this study aims to analyze the impact of dedicated severe accident mitigation/management actions on the results of interest through the established modeling approach. As the modeling and simulation tools for the analysis, MELCOR2.2 and MAAP5.05, were used to compare the analysis results of interest such as the integrities of the reactor coolant system (RCS) pressure boundary and the reactor/containment building, the generation of flammable gases (such as hydrogen and carbon monoxide), and the environmental release of the FP cesium. As a key finding, this study demonstrated that both codes similarly predicted the environmental release of the FP cesium as well as the general trend of the evolution of the severe accident sequences and plant responses. Of particular interest was the substantial difference in predicting the generation of the flammable gases by the two codes. Relevant insights are summarized in terms of particular points of interest.

MODELING LOFW IN A PWR USING MELCOR

The Probabilistic Safety Assessment (PSA) is part of a Nuclear Power Plant (NPP) licensing process. It considers the elaboration and updating of probabilistic models that estimate the risk associated to the operation, allowing the risk monitoring from the design to the plant decommissioning, for both operational as regulatory activities. The PSA identifies those components or plant systems whose unavailability contributes significantly to the Core Damage Frequency (CDF) and to the Large Early Release Frequency (LERF) of radioactive material. Based on the PSA Level 1 results for a reference plant under design, the Analysis, Evaluating and Risk Management Laboratory (LabRisco), located in the University of São Paulo (USP), Brazil, started the analytical investigation of severe accident phenomena using the US Nuclear Regulatory Commission (NRC) MELCOR2.2 code – focusing on the qualification of a group of specialists who will subsidize a PSA Level 2 for the same plant. This PSA Level 1 shows that the accident with large CDF contribution is the Loss of Feed Water Accident (LOFW). Therefore, the initial objective of the investigation was to model the progression of severe accidents during a LOFW for the reference Pressurized Water Reactor (PWR) and to analyze the response of the plant under these accident scenarios. During the course of the hypothetical LOFW in the reference plant, hydrogen was generated – by a reaction between the high temperature steam water and the fuel-cladding inside the reactor pressure vessel (RPV) but not representing a serious threat to the RPV integrity.

Estimation of fission product source terms for the SGTR accident of a reference PWR plant using MELCOR and MAAP5

A steam generator tube rupture (SGTR) is a type of containment bypass accidents that if core damage is not prevented during this accident progression, has the potential to lead to a significant release of fission products (FPs) to the environment through the faulted steam generator (SG) or other possible paths in the secondary system. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during an SGTR accident of a reference pressurized water reactor (PWR) plant, and also investigates the impact of dedicated accident mitigation actions on the results of interest through the established modeling approach. To consider the integral responses of the reactor, SGs, and containment to a severe accident initiated by SGTR, detailed modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. As the modeling and simulation tools for the SGTR accident analysis, MELCOR 2.2 and MAAP5 were employed to compare the results of interest. The results of the analysis demonstrate that the release of FPs into the environment could be estimated more realistically as a result of the best practice modeling for the SGTR accident and also it could be greatly decreased through dedicated mitigation actions. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

Estimation of fission product source terms for the ISLOCA of a reference PWR plant using MELCOR and MAAP5

An interfacing systems loss of coolant accident (ISLOCA) is a type of containment bypass accident that may lead to a significant release of fission products (FPs) to the environment. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during the ISLOCA of a reference pressurized water reactor (PWR) plant and investigates the influence of a dedicated accident mitigation action on the results of interest through the established modeling approach. In addition to this, the key plant features and phenomenological mechanisms that might have a significant influence on the magnitude and timing of FP releases to the environment during an ISLOCA were examined through a couple of sensitivity analyses. To consider the integral responses of the reactor, containment, and auxiliary building to a severe accident initiated by an ISLOCA, detailed modeling was performed for a broad spectrum of relevant phenomena, plant systems, and structures that promote the deposition and retention of FPs along the flow path to the environment. As the modeling and simulation tools for the ISLOCA analysis, MELCOR 2.2 and MAAP5.04 were employed to compare the results of interest. The results of the analysis shows that the present best practice modeling approach proposed here has the potential to significantly decrease the release of FPs into the environment in the event of an ISLOCA. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

An assessment of radiological source terms for loss of cooling and coolant accidents of a PWR spent fuel pool

This paper presents the MELCOR code simulation study regarding how a dedicated severe accident mitigation (SAM) measure the release of radiological FPs to the environment in the spent fuel pool (SFP) of a reference pressurized water reactor (PWR) plant. To draw up relevant insights, a series of accident scenarios have been investigated, taking into account a combination of two initiating events (loss-of-cooling and loss-of-coolant accidents), two SFP operating modes (normal and refueling), and three external water makeup strategies through an external cooling water injection as one of dedicated SAM strategies for a long-term cooling of the SFP. For the analysis, relevant code modeling has been made for a broad spectrum of severe accident-related phenomena being expected during the SFP accidents. As a result, the evolution of relevant physical phenomena has been investigated and discussed in light of the release of FPs, and several points of uncertain issues, which need to address more carefully (a) either when implementing a cross-comparison for different code predictions (such as MELCOR and MAAP) for SFP accident progression or (b) when assessing radiological risk and/or developing dedicated SAM strategies for the SFP of PWR plant, have been drawn up.

Unmitigated severe accident analysis for a PWR using MELCOR

After the two major nuclear accidents in history considered to be beyond-design-basis accidents (BDBA) – the Chernobyl Reactor Four explosion in Ukraine (1986) and the Fukushima Daiichi accident in Japan (2011), the United States Nuclear Regulatory Commission and other countries have included in the Final Safety Analysis Report (FSAR), a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), with significant damage to the reactor core. The Safety Analysis Report (SAR) is the most important document used by a regulatory body to assess the adequacy of plant safety at all stages of the life time of a Nuclear Power Plant (NPP), and presents the licensing basis of a plant.This study aims at comparing the progression of severe accidents during a Total Loss of Feed Water Accident (TLOFW) and a TLOFW combined with An Anticipated Transient Without SCRAM (ATWS) reactivity accident for a Pressurized Water Reactor (PWR) type NPP. This paper takes into account the results of the PSA Level 1 for the reference NPP. The calculations were performed with the deterministic computer code MELCOR version 2.2 used to analyze severe accident progression in NPP with a Light Water Reactor (LWR). The paper presents a hydrogen risk assessment – the assessment is deterministic, s a prerequisite before performing PSA Level 2. The hydrogen source is determined, too. Saphiro's diagram is applied to investigate the containment atmosphere condition (inert or combustible) for two scenarios. The main results of the study reveal that the containment integrity of the reference PWR NPP when both accidents occurred is maintained. However, the core is severely damaged. A future work will investigate mitigation strategies to limit the hydrogen risk during accident.

Evaluation of Discharged Fuel in Preproposed Breed-and-Burn Reactors from Proliferation, Decay Heat, and Radiotoxicity Aspects

The purpose of this technical note is to evaluate the discharged fuel of breed-and-burn (B&B) reactors. The discharged burnup in a B&B core can be high, and there is a concern that as decay heat increases, handling after a shutdown might be difficult. Because discharged fuels contain a number of plutonium nuclides, the potential for proliferation is also a concern. Moreover, radiotoxicity levels are an issue for geological disposal. As reference cores, two stationary wave reactor (SWR) cores proposed in our previous studies were used. The SWR is a special type of B&B reactor. Discharged fuels of the two SWR cores were evaluated by comparing them to a pressurized water reactor (PWR) and a fast breeder reactor. The discharged fuels of both SWR cores were not significantly worse than the reference PWR, even though the burnup was about 2.6 to 7.0 times higher.

The impact of emergency operating safety procedures on mitigation the nuclear thermal power plant severe accident

The most important public concerns is the safety of nuclear power reactors. It is extremely important to protect the public from the nuclear radiation exposure in the undesired case of a nuclear power plant (NPP) accident. The proper performance of the safety systems of nuclear power plants is one of the most highly important concerns which enhances the use of nuclear energy options. This research paper focuses on the response and performance of the emergency safety systems of a pressurized water reactor (PWR) during loss of coolant accident (LOCA) for a selected reference PWR. The simulation depicts the reactor behavior during hot leg loss of coolant accident event and the leakage area is 200 cm2. This was accomplished by modeling the PWR and the passive safety systems employed using PCTran code which is designed to describe the emergency core cooling system (ECCS) response and core behavior under severe accident conditions. The ECCSs are designed to cool down the reactor core and to prevent core damage. Results of the core temperature distribution, reactor coolant inlet and outlet temperatures, reactor coolant mass flow rate, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the PCTran model is capable of contributing to the process of safety analysis of the NPP.

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

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

Economic analysis of grid and wire wrap supported hydride and oxide fueled pressurized water reactors

An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45 wt.%)–ZrH 1.6 hydride and UO 2 oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing the COE is the remaining plant life, beginning at the time of retrofit. The goals of the analysis are twofold: (1) comparing the economic performance of UO 2 and U–ZrH 1.6 fuels for a range of retrofitted geometries supported by grid and wire wrap spacers; and (2) investigating the potential economic benefits for nuclear utilities considering retrofitting new fuels and/or geometries into existing PWR pressure vessels. Fuel cycle, operations and maintenance (O & M), and capital costs are considered. The economic performance of U–ZrH 1.6 and UO 2 fuels is found to be similar, with UO 2 fueled designs providing a slight advantage when supported by grid spacers, and U–ZrH 1.6 providing a slight advantage when supported by wire wrap spacers. These small differences in cost, however, are within the bounds of uncertainty of this study and are not believed to provide a strong economic argument for the use of one fuel type over the other. To demonstrate the potential economic benefits of retrofitted designs to nuclear utilities, two different comparisons are made. The first compares the COE for retrofitted designs with the COE for a reference PWR, assumed to have operated long enough to recuperate its initial capital investment. The costs for this reference PWR reflect the “do-nothing” case for current plant owners whose primary expenditures are fuel cycle and O & M costs. The second comparison introduces a different reference PWR that includes the costs to operate an existing unit and the cost to purchase power from a newly constructed PWR, for comparison with retrofitted designs which offer increased power relative to existing commercial PWRs. For the first comparison, no grid supported designs and only one wire wrap supported design (i.e., U–ZrH 1.6 Stretch Case) provide a lower levelized unit cost of electricity than the reference “do-nothing” PWR. The primary cause of this conclusion is the capital costs incurred by retrofitted designs to change the core geometry and, for many designs, to upgrade primary and secondary loop components for operation at higher power than the reference PWR. The reference “do-nothing” PWR cost in this first comparison includes only operations and maintenance as well as fuel cycle costs but does not include a capital component. For the second comparison, significant cost savings are demonstrated for both grid (15–19% savings) and wire wrap (30–40% savings) supported designs using U–ZrH 1.6 and UO 2 fuels. These cost savings are enabled by enhancing the pumping capacity of the primary system and, for wire wrap supported designs, by taking advantage of enhanced critical heat flux performance. The optimal geometry for retrofitted UO 2 and U–ZrH 1.6 fueled PWR cores supported by grid spacers is D rod = 6.5 mm and P/ D = 1.39. The cost savings over the second case reference PWR are ∼19 and 15%, respectively. The cost savings for retrofitted PWRs that incorporate wire wrap spacing are even larger because of operation at even higher power. Cost savings over the reference PWR range between 30 and 40% for the U–ZrH 1.6 and UO 2 Achievable and Stretch Cases. The optimal geometries for the U–ZrH 1.6 Achievable and Stretch Cases are D rod = 8.08 mm, P/ D = 1.41 and D rod = 8.71 mm, P/ D = 1.39, respectively. The optimal geometries for the UO 2 Achievable and Stretch Cases are D rod = 7.13 mm, P/ D = 1.42 and D rod = 9.34 mm, P/ D = 1.27, respectively. Utilities seeking to meet rising demand by expanding capacity may therefore strongly benefit from retrofitting existing PWRs with either U–ZrH 1.6 or UO 2 fueled designs. These new designs have different geometries than are currently used by commercial plants. A conclusion on which fuel type to use, however, could not be reached in this analysis as both offer similar economic performance.

Recycle of Transuranium Elements in Light Water Reactors for Reduction of Geological Storage Requirements

This paper presents results of a feasibility study aimed at developing a zero-transuranic-discharge fuel cycle based on the U-Th-TRU ternary cycle. The design objective is to find a fuel composition (mixture of thorium, enriched uranium, and recycled transuranic components) and fuel management strategy resulting in an equilibrium charge-discharge mass flow. In such a fuel cycle scheme, the quantity and isotopic vector of the transuranium (TRU) component is identical at the charge and discharge time points, thus allowing the whole amount of the TRU at the end of the fuel irradiation period to be separated and reloaded into the following cycle. The TRU reprocessing activity losses are the only waste stream that will require permanent geological storage, virtually eliminating the long-term radiological waste of the commercial nuclear fuel cycle.A detailed three-dimensional full pressurized water reactor (PWR) core model was used to analyze the proposed fuel composition and management strategy. The results demonstrate the neutronic feasibility of the fuel cycle with zero-TRU discharge. The amount of TRU and enriched uranium loaded reach equilibrium after about four TRU recycles. The reactivity coefficients were found to be within a range typical for a reference PWR core. The soluble boron worth is reduced by a factor of ~2 from a typical PWR value. Nevertheless, the results indicate the feasibility of an 18-month fuel cycle design with an acceptable beginning-of-cycle soluble boron concentration even without application of burnable poisons.

Containment behaviour in the event of core melt with gaseous and aerosol releases (CONGA)

The CONGA project concentrated on theoretical and experimental studies investigating the behaviour of advanced light water reactor containments containing passive containment heat removal systems and catalytic recombiners expected to be effectively operational during a hypothetical severe accident involving large quantities of aerosol particles and noncondensable gases. The central point of interest was the investigation of the effect of aerosol deposition on the condensation heat transfer of specially designed finned-type heat exchangers (HX) as well as the recombination efficiency of catalytic recombiners. A conceptual double-wall Italian PWR design and a SWR1000 design from Siemens were considered specifically as the reference Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) designs. An assessment of selected accident scenarios was performed in order to define the range of boundary conditions necessary to perform the experimental studies of the other work packages. Experimental investigations indicated that aerosol deposition accounted for up to 37% loss in the heat removal capacity of the two-tube-layer BWR HX units. However, no significant heat transfer degradation could be observed for the PWR HX units. These results can be attributed to the important differences in the designs and operating conditions of the two units. The tests to study the effect of hydrogen (simulated by helium) on the heat transfer rate for heat exchanger units designed for BWR and PWR applications indicated a degradation less than 30% under various conditions. This was found to be acceptable within the over capacity designed for the heat exchangers or containment characteristics. The tests performed to study the long-term aerosol behaviour in the pressure suppression chamber of the current operating BWRs indicated that the water pool scrubs the aerosol particles effectively and reduces the ultimate aerosol load expected on the off-gas system. The efficiency of the catalytic recombiner system designed by Siemens for the off-gas system was found to be insensitive to the aerosol deposited in the recombiner. A computation code, HTCFOUL, was developed to predict the heat transfer rate of a finned-type heat exchanger subjected to a steam–noncondensable gas mixture containing airborne aerosol particles. The model predicts the non-aerosol part of two tests within a variation of 26% and the aerosol part within 32%.

00/02548 Global rupture of a nuclear reactor lower head with elevated temperature due to severe accidents

The CONGA project concentrated on theoretical and experimental studies investigating the behaviour of advanced light water reactor containments containing passive containment heat removal systems and catalytic recombiners expected to be effectively operational during a hypothetical severe accident involving large quantities of aerosol particles and noncondensable gases. The central point of interest was the investigation of the effect of aerosol deposition on the condensation heat transfer of specially designed finned-type heat exchangers (HX) as well as the recombination efficiency of catalytic recombiners. A conceptual double-wall Italian PWR design and a SWR1000 design from Siemens were considered specifically as the reference Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) designs. An assessment of selected accident scenarios was performed in order to define the range of boundary conditions necessary to perform the experimental studies of the other work packages. Experimental investigations indicated that aerosol deposition accounted for up to 37% loss in the heat removal capacity of the two-tube-layer BWR HX units. However, no significant heat transfer degradation could be observed for the PWR HX units. These results can be attributed to the important differences in the designs and operating conditions of the two units. The tests to study the effect of hydrogen (simulated by helium) on the heat transfer rate for heat exchanger units designed for BWR and PWR applications indicated a degradation less than 30% under various conditions. This was found to be acceptable within the over capacity designed for the heat exchangers or containment characteristics. The tests performed to study the long-term aerosol behaviour in the pressure suppression chamber of the current operating BWRs indicated that the water pool scrubs the aerosol particles effectively and reduces the ultimate aerosol load expected on the off-gas system. The efficiency of the catalytic recombiner system designed by Siemens for the off-gas system was found to be insensitive to the aerosol deposited in the recombiner. A computation code, HTCFOUL, was developed to predict the heat transfer rate of a finned-type heat exchanger subjected to a steam–noncondensable gas mixture containing airborne aerosol particles. The model predicts the non-aerosol part of two tests within a variation of 26% and the aerosol part within 32%.

日本原子力研究所におけるＬＯＣＡ／ＥＣＣＳ実験研究の成果

Japan Atomic Energy Research Institute has investigated loss-of-coolant accident (LOCA)/ emergency core cooling system (ECCS) from 1970. Major results of the LOCA/ECCS research are summarized in this report. ROSA-II program was LOCA/ECCS research for a pressurized water reactor (PWR) and ROSA-III program was for a boiling water reactor (BWR). The both test facilities were scaled at approximately 1/400 of the respective reference PWR and BWR. Large scale reflood test is research on reflood phenomena during a large break LOCA of PWR. The test facility is scaled at approximately 1/20 of the reference PWR and the research is still being continued.

Extended Burnup Fuel Cycle Optimization for Pressurized Water Reactors

A fast, yet accurate, fuel cycle analysis methodology was developed to optimize the various options for in-core nuclear fuel management. The methodology encompasses two major parts, a multicycle point reactor model, PUFLAC, and a reload pattern optimization code called DSPWR. The PUFLAC model provides a convenient and reliable survey ability to explore the various fuel cycle scheme possibilities while DSPWR utilizes a direct search scheme to minimize the core power peaking with consideration given to local power-peaking factor variation. A two-dimensional nodal code used in this direct search scheme was developed for the power distribution calculations and is based on the widely used code, EPRI-NODE-P, with very good agreement obtained. This methodology has been demonstrated by considering an extended burnup three-to-four batch transition cycle analysis using Zion Unit 1 as a reference pressurized water reactor plant with realistic power-peaking constraints. The four-batch scheme can yield an increase in uranium utilization of about 5% and a decrease in fuel cycle costs of about 7%. The transition from a three to four-batch scheme can yield an overall increase in uranium utilization of 2.4% and a decrease in fuel cycle costs of about 4%. The transition fuel-loading patterns optimized by DSPWR satisfy themore » core power-peaking constraint with a 2 to 3% margin at beginning-of-cycle.« less