SIMULATE-3K

During the autumn of 2008, two broken control rod stems were discovered at two Swedish BWRs. One of them is Forsmark 3. The discovery led to shaft replacements and careful continuous observation programs. The Swedish Nuclear Inspectorate has accepted these solutions. Even so, special studies were required in order to investigate the consequences of control rod drop, with a new assumption: any control rod could fall at anytime. Control rod drops belong to the class of severe reactivity insertion accidents. Under the new assumption, the highest rod worth, for a single control rod, can be above 3000 pcm during the control rod withdrawal sequence. These events may lead to pellet and cladding failure. They require the evaluation of the core response with multidimensional, multi-physics models. The purpose of this paper is to show the applicability of SIMULATE-3K, either standalone or coupled with the safety analysis code RELAP5, to this class of severe reactivity insertion accidents. This paper shows that SIMULATE-3K adequately captures the complicated interaction between physical processes in the core, in comparison with the more detailed RELAP5 thermal-hydraulic model, as well as the adequacy for analysis near the pellet and cladding failure limits. This paper discusses the consequences for the core in relation to the inserted reactivity worth and the amount of fuel fragmentation. This paper shows the calculated reactivity components, pin-by-pin temperature, and enthalpy distributions computed by SIMULATE-3K “explicit pin-by-pin conduction” model, as well as the pin failure statistics.

Coupled neutronics-thermal hydraulic codes are used by many utilities, research institutes and regulatory authorities worldwide for performing BWR stability analysis. RAMONA-3 has been established in the industry for quite a long time as a reliable time-domain dynamic code with best performance for predictive calculations. Next generation of codes such as RAMONA-5, SIMULATE-3K and POLCA-T, with advanced two-group neutronics and more detailed plant description and thermal hydraulics models have been introduced. The performance of these codes against the stability measurements performed in cycle 19 at the Swiss nuclear power plant Leibstadt (KKL), a BWR/6 from General Electric, is presented in this paper. Important suppliers of the nuclear industry such as Westinghouse Electric Sweden, AREVA NP Germany, Studsvik Scandpower, Inc. USA, and the Swiss research institute PSI have participated in this work. The validation of calculation methods against the KKL stability measurements was considered important by the various organizations for different reasons. Amongst others, Studsvik Scandpower aimed at filling a gap in the SIMULATE-3K stability benchmark database to include a jet pumps driven plant, AREVA NP had to fulfill fuel licensing requirements, and Westinghouse planned to launch POLCA-T parallel to a validation of RAMONA-5 as a production code. PSI cooperated with KKL in stability issues from the very beginning and introduced the stability test project in the framework of NACUSP, a European consortium that aimed for a better understanding of the BWR stability problem. For that purpose, this validation provides an assessment of advanced stability codes for modern BWR core designs.

A coupling between the TRACE system thermal-hydraulics code and the SIMULATE-3K (S3K) three-dimensional reactor kinetics code has recently been developed in a collaboration between the Paul Scherrer Institut (PSI) and Studsvik. In order to verify the coupling scheme and the coupled code capabilities the NEA/OECD Turbine Trip benchmark was simulated. The core/plant system data were taken from the benchmark specifications while the nuclear data were generated with Studsvik’s lattice code CASMO4 and core analysis code SIMULATE-3. The comparison with the experimental data shows that the TRACE/S3K code reproduces well the main transient parameters, namely, the pressure wave propagation, void collapsing and core power response.

Coupled multiphysics problems solve different physical phenomena with time scales of varying orders of magnitude. These phenomena are coupled in a nonlinear way making it difficult to find an accurate and efficient solution. Many of the present generation of codes for LWR are based on 3-D neutronic nodal methods coupled with first order thermal hydraulic methods. Moreover, the spatial and temporal meshes used to solve each field are different reflecting the scales of each phenomenon.

This paper discusses the effect of the spatial and temporal discretization as well as the effect of different coupling schemes, with different level of implicitness, between the neutronic and core thermal hydraulics in SIMULATE-3K (S3K). S3K is a best estimate code used by many utilities, regulatory authorities and research institutes for the analysis of LWR transients that require the coupling of neutronic, fuel pin, and core hydraulic models. Examples of S3K applications are BWR stability analysis, fast anticipated operational occurrences, with or without scram, and reactivity initiated transients. Three different applications will be discussed in this paper to illustrate the effect of the discretization and coupling methods in multiphysics problems, namely: the NEA PWR rod ejection, the Ringhals-1 BWR stability, and the Peach Bottom turbine trip benchmarks.

The best-estimate coupled neutronic/thermal hydraulics code, SIMULATE-3K (S3K), is used by many utilities, research institutes, and regulatory authorities in Europe for performing BWR stability analysis. Analysis of many measured BWR stability tests (often performed in European BWRs) provides the basis for the validation for stability parameter (decay ratio and natural frequency) calculations with S3K. This paper summarizes part of the extensive validation data base for the code, and discusses the influence of fuel pin model parameters on the stability results.

The best-estimate neutron kinetics code, SIMULATE-3K, and the best-estimate nuclear systems analysis code, RELAP5-3D, have been coupled to provide a best- estimate coupled code system for performing plant transient calculations with reactivity feedback from a detailed core model. The coupling of the two well known codes provides for a robust and flexible system capable of analyzing current and proposed Light Water Reactor designs. Comparisons to plant data for two transients and a calculation of a Main Steam Isolation Valve Closure with failure to SCRAM transient are presented.

Several organizations, utilities, fuel suppliers and code developers are performing cycle-specific transient predictions for BWRs. The accuracy of the prediction depends on code approximations and code-specific validation results. The penalty that has to be taken in safety analysis depends on the extent and success of the validation method. This paper presents applications of SIMULATE-3K (S3K) to operational transient calculations and results of validation efforts in the Finnish Olkiluoto 1 and 2 plants. The validation models actual events that have occurred at Olkiluoto and includes a fast pressurization and a fast flow reduction that are typical of internal-pump BWRs. Furthermore, the paper discusses the capability that offers S3K to evaluate Operating Limit Minimum CPR (OLMCPR) directly based on 3D transient methods, without the approximations used in traditional 1D evaluations.

The paper describes the application of the SIMULATE-3K code to a class of fast operational BWR transients, that are typically analyzed as part of the core reload design licensing process. The models and methods of the code are presented. Validation results are shown for two recorded fast transient events in the Olkiluoto-1 and –2 reactors. It is concluded that the code adequately captures the complicated interaction between physical processes in the reactor as well as the essential reactor protection and control systems, which qualifies it for applications to this class of fast transients.

The best-estimate neutron kinetics code, SIMULATE-3K and the best-estimate nuclear systems analysis code, RELAP5-3D, have been interfaced to provide a best-estimate coupled code system for performing plant transient calculations with reactivity feedback from a detailed core model. New reactor and fuel designs require more detailed methods for assessing the behavior of the core and NSSS during ATWS, ejected or dropped control rods, and steam-line breaks.

SIMULATE-3K (S3K) is a best-estimate nodal reactor analysis tool that employs advanced core neutronics coupled with detailed thermal-hydraulic channel models. Faithful modeling of assembly-by- assembly neutronic and thermal-hydraulic effects, including assembly pin power reconstruction, permits application of S3K to a wide class of LWR core transients. Utility licensing approval from the U.S. NRC has been obtained for reactivity insertion accidents (RIA) and S3K has been employed in support of dynamic rod worth measurements (DRWM).

Recently, there has been considerable focus on the licensing basis for RIA limits, which impacts both pressurized water reactor (PWR) and boiling water reactor (BWR) RIA analysis. In Europe and Japan, exposure-dependent RIA fuel enthalpy criteria have been adopted. Similarly, a recent NRC recommendation for an interim RIA acceptance criterion has also proposed exposure-dependent limits on fuel enthalpy rise. One consequence of these new criteria is that more detailed analysis of RIA events may be required because fuel pins with highest enthalpy may no longer be the most limiting.

This paper describes recent extensions to the S3K fuel pin modeling to enhance applications to RIA scenarios with exposure-dependent fuel failure criteria (FFC).

The best-estimate neutron kinetics code, SIMULATE-3K, and the best-estimate nuclear systems analysis code, RELAP5-3D, have been interfaced to provide a best- estimate coupled code system for performing plant transient calculations with reactivity feedback from a detailed core model.

SIMULATE-3K is used by many utilities for the prediction and analysis of BWR stability events. In recent years, several enhancements of the SIMULATE-3K channel and vessel thermal-hydraulic model have been made. This paper summarizes the status of the SIMULATE-3K core thermal-hydraulic model and presents results of the validation against the OF64 test section experiments performed in the Frigg loop. The results comprise comparisons of channel pressure drops, void measurements, and stability limits.

SIMULATE-3K is a two-group, advanced nodal reactor analysis transient code. It has been used by many utilities for the prediction and analysis of BWR stability events. In recent years, several enhancements of the SIMULATE-3K channel and vessel thermal-hydraulic model have been made. This paper summarizes the status of the SIMULATE-3K core and vessel models. It presents results for the Ringhals-1 Stability Benchmark and for regional instability in a large BWR. Finally, modeling sensitivities to space and time discretization are discussed.

This paper discusses the model and results for the Peach Bottom 2 Turbine Trip Test 2 using Studsvik Scandpower’s transient code SIMULATE-3K. This transient is currently the subject of a NEA/OCED BWR benchmark. All data pertaining to core, vessel, and scenario were taken from the benchmark specifications. The nuclear data were generated with Studsvik Scandpower’s lattice code CASMO4 and core analysis code SIMULATE-3. Comparisons to measured data, sensitivity to model options and data, as well as results from a more limiting scenario are presented. SIMULATE-3K captures well the parameters of importance in this transient, namely the pressure wave propagation, the void collapse during the pressurization phase, and the resulting power peak.

SIMULATE-3K has been used to investigate the stability margins of Leibstadt. The model has been benchmarked against stability measurements in cycle 19 and used for predictive calculations in cycle 23. The paper presents results of the cycle-19 benchmark and verification of stability exclusion and surveillance regions for the cycle 23 core in the presence of an increasing use of modern fuel bundles with partial-length rods. The close agreement between calculations and measurements helps establish the S3K model as a reliable tool for predictive stability calculations in Leibstadt.