RELAP5-3D code validation in the neutron-dynamic analysis of transient processes taking place in RBMK-1500 reactors

The paper presents an evaluation of RELAP5-3D code suitability to model-specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian complex neutronic-thermal-hydraulic code STEPAN/KOBRA, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN/KOBRA codes, showed reasonable mutual agreement of the calculation results of both codes and their reasonable agreement with the real plant data.     

Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK-1500 reactor main circulation circuit model

Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines.     

Measurements of the isothermal, power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Modelling of the spatial distribution of the induced activities in the RBMK-1500 reactor graphite blocks and rings/sleeves

There is one nuclear power plant (NPP) in Lithuania – the Ignalina NPP – which is under decommissioning now. The Ignalina NPP has two units with RBMK-1500 reactors, which are the most powerful and the most advanced versions of RBMK-type reactor design. Unit 1 of the Ignalina NPP was shut down at the end of 2004 and Unit 2 was shut down at the end of 2009. RBMK is a water-cooled graphite-moderated channel-type power reactor and the decommissioning of these reactors faces specific challenges for proper characterisation and disposal of irradiated reactor graphite.Apart from radiological inventory, the spatial distribution of radionuclides in the reactor graphite is also very important because it could indicate the possibilities for decontamination/treatment of the irradiated graphite. This is important for consideration of the near surface disposal option for irradiated graphite, as without treatment it usually does not meet the waste acceptance criteria.Based on that, the work presented in this paper is focused on the modelling of the induced activity spatial distribution in the Ignalina NPP RBMK-1500 reactor graphite components: blocks and rings/sleeves. The modelling was performed with MCNP and SCALE computer codes and consisted of two mains stages: modelling of the neutron flux in the reactor graphite components, and then modelling of the neutron activation in them using the already modelled neutron flux. In such a way, the spatial induced activity distribution in the analysed reactor components was obtained. Modelling results show that the thermal neutron flux is more intensive in the outer radial regions of the graphite components and this, in general, results in higher induced activities there.     

Development and validation of a transition boiling model for RELAP5/MOD3 reflood simulation

The current version of the RELAP5/MOD3.1 code significantly underpredicts the transition boiling heat transfer during reflooding of hot fuel rods. In order to extend the code’s range of application for LOCA and degraded core analyses, a new transition boiling model has been developed, assessed and implemented. The model is based entirely on local state variables calculated by the code (wall and fluid temperatures, pressure, void fraction, mass flux and static quality) and does not rely on other history parameters, such as quench position or CHF and minimum film boiling temperatures. A number of separate-effect and bundle experiments are analyzed with the modified version of the code, and the predictions are compared with the ones obtained by the current version and with available experimental data. In all cases, the predictions of the improved model better fit the measured data. The shape of the new temperature curves is more physically and conceptually sound than the one calculated by the current version of the code.     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.     

On the use of RELAP5-3D as a subchannel analysis code

The capabilities of the RELAP5-3D code to perform subchannel analyses in sodium-cooled fuel assemblies were evaluated. The motivation was the desire to analyze fuel assemblies with traditional (solid pins) as well as non-traditional (e.g., annular pins with internal cooling, bottle-shape) geometries. Since no current subchannel codes can handle such fuel assembly designs, a new flexible RELAP5-based subchannel model was developed. It was shown that subchannel analysis of sodium-cooled fuel assemblies is indeed possible through the use of control variables in RELAP5. The subchannel model performance was then verified and validated in code-to-code and code-to-experiment analyses, respectively. First, the model was compared to the SUPERENERGY II code for solid fuel pins in a conventional hexagonal lattice. It was shown that the temperature predictions from the two codes agreed within 2% (<3.5 °C). Second, the model was applied to the Oak Ridge 19-pin test, and it was found that the measured outlet temperature distribution could be predicted with a maximum error of 8% (<7 °C). Furthermore, the use of semicircular ribs on the duct wall to flatten the temperature distribution in a traditional hexagonal assembly was explored by means of the newly developed RELAP5-3D subchannel model; the results are reported here as an example of the model capabilities.     

钠冷快堆中熔融池模型的建立与验证

快堆在超设计基准事故下运行时,会导致钠沸腾和干涸,如果不能及时停堆,接着就会产生燃料元件的熔化坍塌,在组件盒下部形成熔融池.为了对熔融池给出合理的安全分析,采用机理建模的方法,建立了完整的熔融池模型,并在法国的SCARABEE系列实验中的BF1三种功率的实验上进行了验证,和实验吻合较好,通过和所验证过的GEYSER及BF2等实验模型进行比较,得出了有关熔融池机理的相关结论.通过排热和温升等相关数据的比较,对熔融池向外的排热形式给出了合理分析,并得出了相关结论.     

Assessment of RELAP5 point kinetic model against reactivity insertion transient in the IAEA 10 MW MTR research reactor

The current study emphasizes an aspect related to the assessment of a model embedded in a computer code. The study concerns more particularly the point neutron kinetics model of the RELAP5/Mod3 code which is worldwide used. The model is assessed against positive reactivity insertion transient taking into account calculations involving thermal-hydraulic feedback as well as transients with no feedback effects. It was concluded that the RELAP5 point kinetics model provides unphysical power evolution trends due most probably to a bug during the programming process.     

Comparative RELAP5-3D analysis in support of the NPP DBA analysis in Ukraine

In the Ukrainian in-depth safety assessment (ISA) projects the computer code RELAP5/Mod3.2 with point kinetics approximation is being used in the deterministic safety analysis of VVERs. It is generally accepted that the use of this approximation, with the proper modeling assumptions, results in conservative results. However, only coupled three-dimensional codes are capable to estimate the real localized feedback effects for such VVER specific transients as control rod ejection or main steam line break. Some results of the comparative RELAP5-3D analysis for the scenarios, that present strong local reactivity effects, are discussed in this paper. The goal of this RELAP5-3D analysis is to examine the differences in results obtained by the three-dimensional approach and the methodology that was used in Ukrainian ISA projects.     

Investigation of the influence of the sodium void effect of reactivity on the technical-economic performance and safety of an advanced fast reactor

Measures to decrease the sodium void effect of reactivity and the influence of this effect on the technical-economic performance and the safety of BN-1200 are analyzed. Three variants of the core structure differing by the structural implementation and dimensions are examined. It is shown that a flattened core with a sodium cavity, replacing the top end screen, gives self-protection of the reactor with respect to unanticipated accidents. The elimination of the sodium cavity and an increase of the core height result in degradation of the self-protection properties but at the same time improve the technical-economic properties of the reactor. The possibilities for optimizing the construction of the reactor from the standpoint of reaching a compromise between safety properties and the technical-economic characteristics are discussed.     

RELAP5-3D modeling of PWR steam generator condensation experiments at the Oregon State University APEX facility

This report describes modeling using RELAP5-3D of a series of six steam generator U-tube steam condensation (without non-condensable gas) tests conducted at the Oregon State University Advanced Plant Experiment Test Facility from 2005 through 2007. These tests were designed to evaluate steam condensation rates in a scaled pressurized water reactor steam generator at various primary and secondary side pressures and inlet steam mass flow rates. Comparisons between the experimental data and the RELAP5-3D model results are made to quantify the effectiveness of RELAP5-3D in handling steam condensation in U-tube steam generators. RELAP5-3D tends to over predict the condensation rate and heat transfer coefficient when compared against the experimental data when the code uses the laminar Nusselt correlation to determine the heat transfer coefficient. When RELAP5-3D results are used with the Shah correlation the comparison between the heat transfer coefficients is much improved.     

快堆中钠环境下熔融物传播模型的建立与验证

快堆在超设计基准事故下运行,会导致燃料元件的熔化,并很快形成燃料和钢的混合沸腾池,池壁破损之后,高温熔融物将向组件盒间隙以及相邻组件中传播,使事故进一步升级.为了弄清熔融物传播的机理,采用了机理建模的方法,分别建立了导热冻结和整体冻结两种机理模型,在英国的SES实验及法国的PV-A实验上进行了验证,得出了有关冻结机理和钠流动的相关结论,通过对传播的相关参数进行比较分析.得出了熔融物传播的相关结论.     

Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its’ own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.     

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF₂, LiF, ZrF₄ and Li₂BeF₄ eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.     

Reactivity temperature coefficients for the HEU and LEU fuel of the MNSR reactor

Analysis of the Reactivity Temperature Coefficients of the Miniature Neutron Source Reactor (MNSR) for normal and accidental conditions (above 45 °C) using HEU-UAl₄ and the LEU: U₃Si, U₃Si₂ and U9Mo fuel were carried out in this paper. The Fuel Temperature Coefficient (FTC), Moderator Temperature Coefficient (MTC), and Moderator Density Coefficient (MDC) were calculated using the GETERA code. The contribution of each isotope presented in the fuel cell was calculated for the temperature range of 20 °C–100 °C at the beginning of the core life. The average values of the FTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were found to be: −2.23E-03, −1.85E-02, −1.96E-02, −1.85E-02 mk/°C respectively. The average values of the MTC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −8.91E-03, −1.24E-04, −4.70E-03, 2.10E-03 mk/°C respectively. Finally, the average values of the MDC for the UAl₄, U₃Si, U₃Si₂ and U9Mo were observed to be: −2.06E-01, −2.03E-01, −2.04E-01, −2.03E-01 mk/°C respectively. It's found also that the dominant reactivity coefficient for all types of fuel is the MDC.     

Analysis of ABWR FW pump performance during power ascending and verification of an operational strategy to jump pump speed across the critical region with RELAP5-3D

The fourth nuclear power plant in Taiwan is an advanced boiling water reactor (ABWR) and is scheduled to be in commercial operation in late 2009. However, it is highly suspected by the reactor vendor that the turbine-driven reactor feedwater pumps (TDRFPs) are over designed. Consequently, the critical speed of TDRFP is likely to be encountered during power ascending. Besides, the original design speed of TDRFP also has to be reduced during normal operation by using oversized TDRFPs. Therefore, the design of FW control system needs to be accordingly revised based on a proper pump speed versus demand curve. To avoid unnecessary effort during pre-operation and/or power tests, a RELAP5-3D based plant integral model covering both reactor system and BOP systems was applied to simulate and analyze the behavior of the FW system during power ascending. By using the most advanced simulation technique, the performance of TDRFPs during power ascending was calculated, and the low power interval correspondent to the range of critical speed (2400–2800 rpm) was identified. Moreover, an operational strategy proposed by plant operators to jump TDRFP across the critical speed range during power ascending was also quantitatively verified. It was found that increasing the bypass flow of either condensate pump or condensate booster pump is the most efficient and practical approach to jump the TDRFP across the critical speed interval. The successful application of the RELAP5 for the entire BOP simulation indicates that the advanced RELAP5 can extend its traditional reactor safety analysis to the entire power conversion system simulation and analysis.     

Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.