Nuclear Power Engineering Corporation Research Articles

Verification and validation of dynamic response simulation codes for BWR fuel assemblies under seismic loading

The purpose of this study is to verify and validate two independently developed simulation codes for investigating dynamic response behaviors of fuel assemblies in a boiling water reactor (BWR) under seismic loading. The core of BWR consists of several hundreds of fuel assemblies. They are supported with both top guide and fuel support and are surrounded by coolant water. It is important to grasp their dynamic response behaviors under seismic loading for securing the structural integrity of the fuel assembly itself as well as for assessing control rod scrammability. For this evaluation of dynamic response behaviors, reliable simulation codes are required. In this study, we employ two different numerical simulation methods of acoustic fluid-structure interaction (AFSI) developed by the present authors independently. The one is a three-dimensional parallel finite element method for AFSI problems with solid elements based on a partitioned coupling approach, while the other is a finite element method of beam elements for fuel assemblies combining added mass matrix, which represents coupled inertia effects caused by coolant water. Both methods are implemented and applied to a problem of 368 fuel assemblies for verification and validation. The problem was set up based on the demonstration test performed by Nuclear Power Engineering Corporation in 1986. Both simulation results agreed well with each other, and the simulated results also agreed well with the experimental ones. In addition, we have discussed seismic response behaviors of fuel assemblies, which were not shown in the demonstration test. Finally, we conclude that the both developed simulation codes based on the proposed methods are powerful tools to grasp the behavior of huge number of fuel assemblies of BWR under seismic loading and to improve the seismic safety design of BWR core.

Validation of CTF pressure drop and void predictions for the NUPEC BWR database

Abstract A quantified validation of CTF pressure drop, equilibrium quality, and void fraction predictions is performed using the Japanese Nuclear Power Engineering Corporation boiling water reactor database. Four quantities of interest are compared between the experiments and the code predictions: pressure drop, average exit equilibrium quality, average exit void fraction, and subchannel exit void fraction. These four quantities of interest have root-mean-squared errors of 0.124, 0.005, 0.065, and 0.089, respectively. Pressure drop predictions are generally underpredicted for single phase cases and overpredicted for two phase cases. The equilibrium quality predictions are mostly within 0.01 of the designed experimental values, which indicates proper energy conservation. The void fraction results tend to be overpredicted by about 0.06, which is attributed to the interfacial modeling in the code. By splitting the subchannels into different groups, it is shown that those near unheated surfaces are the least accurately modeled, especially for cases that have a high concentration of unheated surfaces. The results are consistent with past validation analyses using a variety of subchannel codes. Unlike past studies with CTF, this work incorporates all 69 pressure drop experiments and 392 steady state void experiments from the database. A new method for quantification of asymmetries is proposed and applied, but the interpretation of the results depends on the definition of the measurement error.

An advanced new fully implicit numerical method for two-phase flow subchannel analysis based on the Drift Flux Model

In this paper, a new and efficient fully implicit numerical model is introduced using the Drift Flux Model (DFM) for simulation of the reactor core thermal-hydraulics using subchannel analysis. Although several fully implicit models have been developed to simulate two-phase flow in a single channel, but development of an efficient model for more realistic conditions, i.e. reactor core, would be useful. Based on the benefits of the Newton method, a procedure for the accurate approximation of the inverse of the Jacobian matrix and a fully implicit numerical scheme is developed. To benchmark the present model, a well-scaled 8×8 rod bundle was simulated and fifteen steady-state test series and two transient cases were selected to analyses the subchannel grade void distributions in NUPEC (Nuclear Power Engineering Corporation) 8×8 rod bundle test facility. The steady-state void distributions predicted by the model are in agreement with the measured data for a wide range of thermal-hydraulic parameters investigated. Transient calculations were also performed with different time scales and it is concluded that the model is not subject to any time step restrictions.

Critical power prediction by CATHARE2 of the OECD/NRC BFBT benchmark

This paper presents an application of the French best estimate thermal-hydraulic code CATHARE 2 to calculate the critical power and departure from nucleate boiling (DNB) exercises of the International OECD/NRC BWR Fuel Bundle Test (BFBT) benchmark. The assessment activity is performed comparing the code calculation results with available in the framework of the benchmark experimental data from Japanese Nuclear Power Engineering Corporation (NUPEC).Two-phase flow calculations on prediction of the critical power have been carried out both in steady state and transient cases, using one-dimensional and three-dimensional modelling. Results of the steady-state critical power tests calculation have shown the ability of CATHARE code to predict reasonably the critical power and its location, using appropriate modelling.

Reducing Uncertainties via Predictive Modeling: FLICA4 Calibration Using BFBT Benchmarks

This work applies the predictive modeling procedure formulated by Cacuci and Ionescu-Bujor [Nucl. Sci. Eng., Vol. 165, p. 18 (2010)] to assimilate experimental data from the international Organisation for Economic Co-operation and Development/U.S. Nuclear Regulatory Commission boiling water reactor full-size fine-mesh bundle test (BFBT) benchmarks to calibrate and reduce systematically and significantly the uncertainties in the predictions of the light water reactor thermal-hydraulic code FLICA4. The BFBT benchmarks were designed by the Nuclear Power Engineering Corporation of Japan for enabling systematic validation of thermal-hydraulic codes by using full-scale experimental data. This work specifically uses BFBT experimental data for the “pump trip for a high-burnup assembly” in the predictive modeling formalism to calibrate parameters and time-dependent boundary conditions (power, mass flow rates, and outlet pressure distributions) in FLICA4, yielding best-estimate predictions of axial void fraction distributions. The resulting uncertainties for the best-estimate time-dependent model parameters and void fraction response distributions are shown to be smaller than the a priori experimental and computed uncertainties, thus demonstrating the successful use of predictive modeling for the large-scale reactor analysis code FLICA4 using BFBT benchmark-grade experiments.

Development of drift-flux model based on 8 × 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 × 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 × 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data.

Benchmarking a sub-channel program based on a drift-flux model with 8 × 8 NUPEC BWR rod bundle

This study deals with the verification of a sub-channel analysis computer program for steady-state and operational transients in boiling water reactors (BWR). Conservation equations based on a flow regime dependent drift-flux model, proper constitutive equations and correlations are assigned to evaluate sub-channel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-size fine-mesh Bundle Tests) facility. Fifteen steady-state test series and two transient cases were selected to analyze the sub-channel void distributions. It was revealed that the steady-state void distributions calculated by the program agreed well with the measured data in the range of thermodynamic qualities from 5% to 25%. The results of the transient calculations were also to experimental data.

A subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark

The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8×8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment.

Overview and Discussion of the OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests

The Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC) benchmark based on the Nuclear Power Engineering Corporation (NUPEC) pressurized water reactor (PWR) subchannel and bundle tests (PSBTs). The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA) of OECD and the Japan Nuclear Energy Safety Organization (JNES), Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs) codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB), under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases.

Best-Estimate Predictions and Model Calibration for Reactor Thermal Hydraulics

This work applies a recently developed best-estimate data assimilation and model calibration methodology to the three-dimensional reactor thermal-hydraulics simulation and design tool FLICA4. The experimental information used for calibrating FLICA4 parameters stems from the international Organisation for Economic Co-operation and Development/Nuclear Regulatory Commission boiling water reactor full-size fine-mesh bundle tests (BFBT) benchmarks, which were designed by the Nuclear Power Engineering Corporation of Japan for enabling systematic validation of simulation tools using full-scale experimental data. The following specific BFBT experiments have been used in this work for calibrating parameters and boundary conditions for FLICA4: (a) axial void fraction distributions and (b) transversal void fraction distributions. The resulting uncertainties for the predicted parameters and distributions of pressure drops and void fractions are shown to be smaller than the a priori experimental and computed uncertainties, thus demonstrating the successful calibration of a large-scale reactor core thermal-hydraulics code using the BFBT benchmark-grade experiments.

Analysis of the MOX critical experiments BASALA by pin-by-pin core analysis method using three-dimensional direct response matrix

An analysis of the MOX critical experiments BASALA was performed to verify the pin-by-pin core analysis method using a three-dimensional direct response matrix. The BASALA experiments simulate full MOX BWR cores, and they were carried out in the EOLE critical facility of the French Atomic Energy Commission (CEA) by the Nuclear Power Engineering Corporation (NUPEC) in collaboration with CEA. The BASALA experimental cores are very heterogeneous because their size is much smaller than that of commercial power plants. The main features of the pin-by-pin core analysis method using the three-dimensional direct response matrix are that the response matrix can reflect the intra-assembly heterogeneous effect, the diffusion approximation is not involved, and the fuel rod fission rate can be directly evaluated. The maximum difference of the critical k-effective values among all nine cores analyzed was about 0.4% Δk. The root mean square differences between the calculated and measured radial fuel rod fission rate distributions in the test assembly of all cores were within 1.8% and nearly comparable to measurement error. The calculated results of the reactivity worth agreed with the measured results within 9%. These good agreements mean that the pin-by-pin core analysis method using the three-dimensional direct response matrix accurately reflects the effects of the intra- and inter-assembly heterogeneities in heterogeneous systems like the BASALA experimental cores.

OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

NUPEC BFBT SUBCHANNEL VOID DISTRIBUTION ANALYSIS USING THE MATRA AND MARS CODES

The subchannel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility were evaluated with the subchannel analysis code MATRA and the system code MARS. Fifteen test series from five different test bundles were selected for an analysis of the steady-state subchannel void distributions. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 % to 25 %. The results of the transient calculations were also similar and were highly feasible. However, the computational aspects of the two codes were clearly different.

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses.

Assessment of integrity of a dual-purpose metal cask after long-term interim storage: seal performance under transport conditions

AbstractSpent fuels generated in nuclear power plants (NPPs) must be stored until they can be reprocessed into new energy sources in Japan. The quantity of spent fuel stored at each NPP site is increasing, and early realisation of a method of interim storage is expected. Dual-purpose metal casks will be used which will not be reopened until they are delivered to a reprocessing plant, in order to minimize radiation exposure to personnel. The Japan Nuclear Energy Safety Organization (JNES) was established on 1 October 2003 with the mission of ensuring public safety from the potential hazards of nuclear energy. JNES have also been conducting a study of dry storage technology for interim storage. The study of verification of metal cask storage technology, the results of which from a storage and subsequent transport safety point of view are presented in this paper, was originally initiated with government funds in 1999 at the Nuclear Power Engineering Corporation, before being transferred to and conducted by J...

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.

Application of Probabilistic Safety Assessment to the Pipe Rupture Incident at Hamaoka Unit-1

On November 7, 2001, a pipe rupture occurred in the steam condensation line of the residual heat removal system at the Hamaoka Unit-1 while operating at rated power, resulting in steam release with radioactivity into the reactor building and the high pressure coolant injection being unavailable. The reactor was manually shut down immediately after the pipe rupture and there was no radioactive release into the environment. Although this incident was not safety significant, the type of rupture was unprecedented. Thus, the regulatory body, the Nuclear and Industrial Safety Agency (NISA), investigated event causes and examined corrective actions for preventing recurrence, and issued the report describing the investigation results, on May 13, 2002. At the NISA's request, the Nuclear Power Engineering Corporation (NUPEC) evaluated the safety significance of the incident and the effectiveness of corrective actions from the risk point of view by carrying out the Probabilistic Safety Assessment (PSA), that is, the estimation of the core damage frequencies (CDFs). The results showed that the incident has no risk significance and that the individual corrective actions would not lead to any significant increase in CDFs. The present study is the first PSA application to the regulatory decision-making process in Japan.

Application of Probabilistic Safety Assessment to the Pipe Rupture Incident at Hamaoka Unit-1

On November 7, 2001, a pipe rupture occurred in the steam condensation line of the residual heat removal system at the Hamaoka Unit-1 while operating at rated power, resulting in steam release with radioactivity into the reactor building and the high pressure coolant injection being unavailable. The reactor was manually shut down immediately after the pipe rupture and there was no radioactive release into the environment. Although this incident was not safety significant, the type of rupture was unprecedented. Thus, the regulatory body, the Nuclear and Industrial Safety Agency (NISA), investigated event causes and examined corrective actions for preventing recurrence, and issued the report describing the investigation results, on May 13, 2002. At the NISA's request, the Nuclear Power Engineering Corporation (NUPEC) evaluated the safety significance of the incident and the effectiveness of corrective actions from the risk point of view by carrying out the Probabilistic Safety Assessment (PSA), that is, the estimation of the core damage frequencies (CDFs). The results showed that the incident has no risk significance and that the individual corrective actions would not lead to any significant increase in CDFs. The present study is the first PSA application to the regulatory decision-making process in Japan.

Post DNB Heat Transfer Experiments for PWR Fuel Assemblies

Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis.

Thermal-Hydraulic Analysis of the Nuclear Power Engineering Corporation Containment Experiments with GOTHIC

GOTHIC version 7.0 was used to model five tests that were conducted in the Nuclear Power Engineering Corporation facility in Japan. The tests involved steam and helium injection into a preheated, spray-moderated, ¼-scale model of a pressurized water reactor dry containment. Comparison of GOTHIC predictions to measured data for pressure, vapor temperatures, structure surface temperatures, and helium concentrations provided the opportunity to evaluate methods for modeling gas dispersion, drop heat and mass transfer, and surface heat transfer.The test facility includes three floors. The lower two floors are partitioned into a variety of rooms that simulate the lower regions of the modeled containment. On the upper floor, rooms that simulate the steam generator enclosures and the pressurizer enclosure extend into the dome, which represents about two-thirds of the total volume of the containment.The GOTHIC model was defined with 30 control volumes using a mix of lumped parameter volumes and subdivided volumes that employ a three-dimensional mesh. Each volume included several thermal conductors to model the various structures. More than 100 flow paths were used to model the hydraulic connections.Comparison of predictions to data showed that enhanced grid resolution in the vicinity of the steam-helium release point served to limit dispersion of the steam-helium plume. The data comparisons also suggested that spray effectiveness was reduced by drop impact with the containment wall and by the high drop concentration. The data comparisons further suggested that the presence of condensation, sprays, splashing, and other wetting mechanisms should be considered to obtain a reasonable estimate of the effect of liquid films on the structure surfaces.