BETA experiments in verification of the Wechsl code: experimental results on the melt-concrete interaction

The BETA experiments are conducted to investigate the melt-concrete interaction in a large-scale melt facility using internally heated simulated core melts. The experimental findings are extrapolated to reactor accident conditions by means of computer codes verified experimentally.The experiments cover a wide range of temperatures and power rates typical of accident conditions. In high-temperature melts, fast downward erosion determines the cavity shape and the very high downward heat transfer causes the temperature of the melt to drop rapidly, even with high internal heating. Crust formation at the interface between the melt and the concrete during the low-temperature interaction allows the gases evolved by the concrete to percolate through the melt, thus establishing an effective gas driven mode of heat transfer. Measurements of gases and aerosols are reported and discussed for silicate and limestone types of concrete.     

Small break loss of coolant accidents: Bottom and side break

A loss of coolant accident (LOCA) in a nuclear reactor can be caused, e.g., by a small break in the primary cooling system. The rate of fluid escaping through such a break will define the time until the core will be uncovered. Therefore the prediction of fluid loss and pressure transient is of major importance to plan for timely action in response to such an event. Stratification of the two phases might be present upstream of the break, thus, the location of the break relative to the vapor-liquid interface and the overall upstream fluid conditions are relevant for the calculation of fluid loss. Experimental results and analyses are presented here for small breaks at the bottom or at the side of a small pressure vessel.It was found that in such a case the onset of the so-called “vapor pull through” is important but swelling at sufficient depressurization rates of the liquid due to flashing is also of significance. It was also discovered that in the bottom break the flow rate is strongly dependent on the break entrance quality of the vapor-liquid mixture. The side break can be treated similarly to the bottom break if the interface level is above the break.The analyses developed on the basis of experimental observations showed reasonable agreement of predicted and measured pressure transients. It was possible to calculate the changing interface level and mixture void fraction history in a way compatible with the behavior observed during the experiments.Even though the experiments were performed at low pressures, this work should help to get a better understanding of physical phenomena occurring in a full scale small break loss of coolant accident.     

The verification basis of the PM-ALPHA code

An overall verification approach for the PM-ALPHA code is presented and implemented. The approach consists of a stepwise testing procedure focused principally on the multifield aspects of the premixing phenomenon. Breakup is treated empirically, but it is shown that, through reasonable choices of the breakup parameters, consistent interpretations of existing integral premixing experiments can be obtained. The present capability is deemed adequate for bounding energetics evaluations.     

Experimental database on fast-reactor thermohydraulics for computational code verification

The results of benchmark experiments on the thermohydraulics of a model fuel assembly and models of targets used in an accelerator-controlled system with a liquid-metal coolant are presented. Calculations performed using the domestic MIF code and foreign codes are compared. The data obtained from experimental studies of the hydrodynamics and heat-exchange in a fast-reactor core cooled by liquid sodium, which were performed primarily at the Physics and Power Engineering Institute over a long period of time, are briefly reviewed. It is shown that the experimental data for use as verification tests need to be systematized, evaluated, and generalized. The analysis shows the degree of completeness of the experimental data and permits making recommendations for performing experiments to fill the existing lacunas.     

基于OpenMC执行码的编译检测及临界基准验证

开源蒙特卡罗程序OpenMC(OpenMonte Carlo code)只提供源代码而没有执行码,在编译OpenMC的过程中发现不同版本的辅助程序与之存在兼容性问题。本文通过分析OpenMPI、Mpich及HDF5各版本辅助程序,对0.6.2版本OpenMC源代码的支持情况进行研究,为正确编译OpenMC执行码给出了直接参考。为进一步验证OpenMC执行码计算临界问题的正确性,选择国际临界安全基准评价实验手册(The International Criticality Safety Benchmark Evaluation Project,ICSBEP)中的96道代表性例题进行基准校验,与通用蒙特卡罗程序的计算结果进行对比并以实验值作为参考。结果表明,OpenMC计算值与实验值及其他程序计算值吻合较好,验证了OpenMC临界计算的可行性和正确性,上述结论将为程序以后的实际应用及完善奠定基础。     

Comparison of ICECO code predictions with flexible vessel experiments

Some comparisons of ICECO code predictions with experimental data concerning transient fluid-structure interaction are given. The test results are taken from flexible vessel experiments conducted by Stanford Research Institute under the direction of Argonne National Laboratory. Two different experiments are considered: one with a rigid core barrel, and one with a flexible core barrel. Both experiments are performed in simple reactor vessels with a well-defined energy source and simple boundary conditions. Correlations of pressures and impulses are made at all available gauge stations. The permanent deformations of the core barrel and the cylindrical vessels are compared with ICECO predictions. The effects of core barrel flexibility on the wave propagation and vessel deformation are also investigated. The agreement between the analysis and experiments is found to be quite good.     

IIST small break LOCA experiments with passive core cooling injection

The purpose of this study is to evaluate the performance of a passive core cooling system (PCCS) with passive injection during the cold-leg small break loss-of-coolant accidents (SBLOCAs) experiments conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. Four tests were performed simulating break sizes of 0.2–2% (approximately corresponding to 1.25–4″ breaks for a referenced nuclear power plant) at cold-leg for assessing the PCCS capability in accident management. The key thermal–hydraulic phenomena to core heat removal for PCCS are observed and discussed. The experimental results show that the PCCS has successfully provided a continuous removal of core heat and a long term core cooling can be reached for all cases of SBLOCA.     

Improvement of HYCA3D code and experimental verification in rectangular geometry

Concerns about the local hydrogen behavior in a nuclear power plant (NPP) containment during severe accidents have increased with the 10CFR50.34(f) regulation after TMI accident. Consequently, investigations on the local hydrogen behaviors under severe accident conditions were required. An analytical model named HYCA3D was developed at Seoul National University (SNU) to predict the thermodynamics and the three dimensional behavior of a hydrogen/steam mixture, within a subdivided containment volume following hydrogen generation during a severe accident in NPPs. In this study, the HYCA3D code was improved with a steam condensation and spray model, and verified with hydrogen mixing experiments executed in a SNU rectangular mixing facility. Helium was used to simulate hydrogen in both the calculations and the experiments. The calculation results show good agreement with the experimental data.     

Comparison of REXCO code predictions with flexible vessel experiments

This paper deals with the REXCO code predictions of the flexible vessel tests. Comparisons are made with the data from experiments which were designed specifically for the purpose of code verification and/or modification. These experiments were performed with a well-defined and calibrated energy source in cylindrical vessel with precision tolerances and simple well-defined boundary conditions. The experimental data can thus be used as reliable test data for validation of computer codes, as well as of the modeling techniques used in the computer analysis. The inputs to the computer analysis are the vessel and core barrel dimensions and boundary conditions, the stress-strain relationships for the vessel and core barrel materials, the equation of state for the coolant, and the pressure-volume relationship of the energy source. The REXCO-predicted wall deformations, pressure loadings, and integrated impulses at various gauge positions are compared with the experimental data. Results of the comparisons are discussed.     

PRTHRUST-J1 code for calculation of blowdown thrust force and its experimental verification

This paper presents an outline of the PRTHRUST-J1 code for calculating blowdown thrust force and gives two numerical examples to show the effectiveness of this code. One numerical example is the problem of saturated steam blowdown. The blowdown thrust forces obtained from the PRTHRUST-J1 code were compared with those of the simplified method of Moody. Fairly good agreement was found between these two results. The other numerical example is the problem of jet discharging tests with stop valve performed in Japan Atomic Energy Research Institute. Analysis was carried out by varying the discharge coefficient. The analytical blowdown thrust force and pressure in the discharging nozzle were compared with experimental results. Qualitative agreement was found between the analytical and experimental results of the blowdown thrust force. Generally speaking, the blowdown thrust forces obtained from the experiment were between the analytical results for discharge coefficients of 1.0 and 0.6.     

Comparison of icepel code predictions with straight flexible pipe experiments

Experimental results on the pressure pulse propagation along plastically deforming pipes are interpreted. The tests are also analyzed by the ICEPEL piping code and the results compared with the experimental measurements of both pressure and circumferential strain histories in the pipe.     

Neutron constants verification by nonstationary experiments with multiplying systems

The arrangement of nonstationary experiments with multiplying systems whose results can be used to verify neutron constants libraries is described. For two multiplying systems with different ratio of plutonium and highly enriched uranium, the experimental results are compared with the computational results obtained using the BAS and ENDF-BVI neutron constants.     

The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis

In September 1988, the United States Nuclear Regulatory Commission issued a revised emergency core cooling system rule for light water reactors that allows, as an option, the use of best estimate plus uncertainty methods in safety analysis. To support the 1988 licensing revision, the United States Nuclear Regulatory Commission and its contractors developed the code scaling, applicability and uncertainty evaluation methodology to demonstrate the feasibility of the best estimate plus uncertainty approach. The phenomena identification and ranking table (PIRT) process, Step 3 in the code scaling, applicability and uncertainty methodology, was originally formulated to support the best estimate plus uncertainty licensing option. Through further development and application, the PIRT process has shown additional utility as a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. The generic PIRT process, including typical and common illustrations from prior applications that promoted further development of the process, are described. Analysis of the results of the prior applications is also described. The analysis results provide information that can help guide future applications of the process in a graded approach based on phenomena relative importance.     

Design,experiments and Relap5 code calculations for the perseo facility

Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories.The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe.ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained.An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses.This paper deals with the experimental and calculated results limited to the Relap5 code.     

Assessment of scaling principles for the simulation of small break LOCA experiments in PWRs

The choice of the scaling laws to be applied for the simulation of nuclear reactor behaviour and, more particularly, the extrapolation of data measured in experimental facilities to real plants, remains an important unresolved issue in nuclear safety.After the analysis of scaling principles adopted in the design of four PWR simulators, the above problem is dealt with in this paper.The definition of a counterpart test and a code analysis, comparing LOFT measured data with calculated trends in the PWR-PUN plant and in LOBI/MODI, LOBI/MOD2 and SEMISCALE facilities, make it possible to check the validity of the criteria utilized in the design of the experimental loops and to reduce uncertainty margins in predicting PWR behaviour.     

Development and verification of an MOC code employing assembly modular ray tracing and efficient acceleration techniques

In this paper, we report the development and verification of a method of characteristics (MOC) code, PEACH, at Shanghai Jiao Tong University. Both the usual flat-source step characteristics (SC) scheme and the linear source (LS) approximation scheme are adopted for the tracking calculation along the neutron trajectory. The assembly-based modular ray tracing (AMRT) technique that possesses a good geometric flexibility and high efficiency is employed, which makes PEACH able to deal with practical LWR assembly and core problems. Moreover, in order to reduce the computational time of the MOC iteration process, both the multi/few-group two-level cell-based coarse mesh finite difference (CMFD) acceleration and the exponential function interpolation technique are used. This results in a significant acceleration. Numerical results for the OECD NEA C5G7 MOX benchmark problem and a 69-group BWR mini-core problem demonstrate that PEACH is accurate and efficient. Numerical results also demonstrate that the LS scheme is more efficient than the SC scheme, taking less time and system memory to generate results of comparable accuracy. In addition, we find that MOC with CMFD acceleration always converges with almost the same number of outer iterations regardless of the physical problem size and the discretization parameters used. This shows an ideal linear relationship between the run time and the geometric size of the problem.