Pressurized Water Reactor Design Research Articles

Integrated calculation method for pressurized water reactor design basis source terms based on linear chain method

Integrated calculation method for pressurized water reactor design basis source terms based on linear chain method

Evaluation of Passive Containment Cooling with an Advanced Water Film Model in a Lumped-Parameter Code

Falling water films have been employed for passive containment cooling in several Generation III pressurized water reactor designs. In this paper, the lumped-parameter (L-P) containment code system COCOSYS with an advanced water film model is applied to evaluate the performance of a passive containment cooling system (PCCS) during accidents. Based on the recent work and with further modification, an integrated water film model is developed. The new model considers different flow regimes of a liquid film as it flows downward and is being evaporated. The integrated model has been adapted to the L-P code and then implemented into COCOSYS. The new model enables the containment code to capture previously neglected phenomena, including the behavior of film breakup due to the reduction in mass; the formation of rivulets; the change in coverage rate and the development of rivulets; the change of velocity distribution as well as film thickness by considering the interfacial shear stress created by countercurrent air on the film surface; the hysteresis of rivulets, i.e., the process of advancing or retreating, involving changes in contact angles; and the influence of waves on the film surface.The new model is validated against existing test results and experimental observations in the authors' recent work and is further modified in this paper taking into account the influence of waves and the processes of rivulet hysteresis. The model is then assessed based on test nodalization, and the expected phenomena are observed. Afterward, the new model is applied to evaluate the performance of PCCS film cooling employed in the AP1000 containment.It is concluded that the original film model tends to underestimate the pressure loads due to the absence of film breakup, rivulet behavior, and shear stress models. The coverage rate, as a new factor captured in the new model, limits the evaporation rate and thus restricts the cooling efficiency of the falling film. The sensitivity analysis reveals that the contact angle and hysteresis phenomenon, which were not previously considered in the code, play significant roles in PCCS film cooling. The advancing contact angle of the rivulets is a decisive factor for the peak pressure, while the retreating contact angle is influential in the later phase of cooling. It can be inferred from the study that the ideal situation for PCCS cooling is that in which the water film is approaching complete dryout at the bottom of the containment. The newly developed liquid film model helps improve the accuracy and reliability of the simulation results.

Neutronics and thermal hydraulics analysis of a small modular reactor

The small modular reactor (SMR) offers many feasible pathways for the construction of more nuclear power plants. A physics model of a near term deployable SMR of the integral pressurized water reactor (IPWR) design is developed. Fuel depletion simulations are performed to optimize the active fuel length, fuel enrichment and core loading pattern in order to achieve a uniform core power distribution. The optimized core can produce 500MW of thermal power with a four year core life-time at a capacity factor of 87%. The core consists of 69 uranium dioxide (UO2) fuel assemblies; 5 assemblies at 4.4at% 235U enrichment and 64 assemblies at 4.95at% 235U enrichment. The active fuel length is 200cm and the core diameter is 194.55cm for an active core height-to-diameter ratio of 1.03. As part of the study the active fuel length is increased to 240cm resulting in an increased capacity factor of 95% at 530MW of thermal power output for an active core height-to-diameter ratio of 1.23. Rod cluster control assemblies (RCCAs) are placed strategically to reduce the overall core power peaking factor to 1.3. Estimated reactor kinetics parameters such as the delayed neutron fraction and mean neutron generation time are typical of existing larger pressurized water reactors (PWRs) from which much of the IPWR based SMR design is derived. This study showed that Doppler, moderator temperature, void and power reactivity coefficients are all negative over the core life-time of four years indicating the possibility of safe reactor operation. A semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin from industry standards on peak fuel and clad surface temperature limits. The critical heat flux (CHF) is calculated and is not exceeded even in 10% overpower conditions. In addition the nucleate boiling ratio (DNBR) is calculated and found to be above 4.8 for the entirety of the active core region. These parameters further engender confidence in the safety of the SMR design.

Boron-free small modular pressurized water reactor design with new burnable absorber

Summary This paper presents a preliminary design of a Small Modular Pressurized Water Reactor (SMPWR) aimed at boron-free operation which can reduce the size of a nuclear power plant (NPP) and the amount of liquid radioactive waste, and also reduce the corrosion issues caused by boric acid in the coolant. The design parameter limits, such as reactivity swing, axial offset (AO), 3D pin peaking factor (Fq), and the required shutdown margin, have been established for the boron-free SMPWR. Furthermore, a new ring type of burnable absorber (R-BA) with Zr–167Er is adopted as a burnable absorber (BA) and HfB2 is adopted as a control rod material to satisfy those design limits without liquid boron. Optimal fuel assemblies (FAs) and loading patterns have been searched through a sensitivity study, and the excess reactivity control capacity and design requirements have been demonstrated to be satisfactory. Copyright © 2016 John Wiley & Sons, Ltd.

Integrity analysis of a reactor pressure vessel subjected to a realistic pressurized thermal shock considering the cooling plume and constraint effects

The fracture mechanic analysis of a reactor pressure vessel subjected to Pressurized Thermal Shock (PTS) loading is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most severe scenario occurs during cold water injection in the cold leg due to a Loss-Of-Coolant Accident (LOCA). Two hypothetical LOCAs are assumed for an adopted reference design of a two-loop pressurized water reactor. Boundary conditions obtained from the RELAP5 code are used as input for three-dimensional computational fluid dynamics which provides the accurate description of the transient including the plume cooling effect. The safety assessment considers the comparison of the stress intensity factor for the deepest point of a surface crack front with the fracture toughness of the material. However, the fracture toughness of the materials is influenced by crack constraints determined by the T-stress. Therefore, safety margin of the reactor pressure vessel should be based on the K-T instead of on the K approach. The existence of a cooling plume affects the stress intensity factor and T-stress. Even if the cooling plume affects the T-stress, it is not translated into an increase of the safety margin. Therefore, in a PTS event with pronounced plume cooling a detailed model of the plume has to be considered in the structural analysis.

2015 Frank Newman Speller Award: Stress Corrosion Cracking and Nuclear Power

In order to provide background for a review of stress corrosion cracking (SCC) in nuclear power plants, a review is first presented of the history of occurrence of SCC in fossil fired power plants and a few related applications. The historical development of SCC in nuclear power plants is then described. The many materials and components that have been affected by SCC over the years are identified, as are the material, stress, and environmental factors involved. A particular focus of this monograph is the question of how it happened that materials with significant susceptibility to SCC were used for so many important structural applications in nuclear power plants. To address this question, the historical development of both pressurized water reactor designs and boiling water reactor designs is reviewed. The review covers factors such as the material types selected for test and use, the operating temperatures used as the designs evolved, and the qualification tests and prior operating experience that were relied upon for selection of materials. Some conclusions as to the factors involved in the choice of materials that turned out to be susceptible to SCC are presented, together with some suggestions for how to minimize such problems in the future.

Comparison of PTS analyses of RPVs based on 3D-CFD and RELAP5

The integrity of a reactor pressure vessel related to pressurized thermal shocks is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most critical scenario occurs during cold water injection through the cold leg due to a Loss-Of-Coolant Accident (LOCA). Due to the difficulties associated with the crack modeling with the three-dimensional finite element method (FEM), simple geometries and crack configurations are usually employed.In the present study, a hypothetical medium break LOCA is assumed in one of the hot legs for an adopted reference design of a two-loop pressurized water reactor. The boundary conditions obtained from RELAP5 calculations are used as input for the three-dimensional computational fluid dynamics simulations in order to provide three-dimensional temperature distribution for the structural mechanics analysis in which submodeling and eXtended FEM (XFEM) are applied. The results from these three-dimensional computations are compared with those from simplified axisymmetric models based on reference data temperatures.

Experimental study on circulation characteristics of secondary passive heat removal system for Chinese pressurized water reactor

Nuclear safety has attracted increasingly global attention. The secondary side natural circulation heat removal system is one of several passive safety systems designed to ensure the safety of the reactor core. The passive heat removal systems are also adopted in the safety system design of Chinese third-generation pressurized water reactor. To complete its design and verify its feasibility, an experimental facility, which is named SPHRS, was built to investigate the heat removal capacity and the stability characteristics of the steam generator secondary side passive heat removal system. The experimental facility consists of a steam generator simulator, an instrumented condenser test section with a secondary pool boiling section, a pump, a water storage tank, and associated piping and instrumentation. The heat removal capacity under varying pressures and characteristics of natural circulation at steady-state as well as system power input drop condition were obtained. The results show that the SPHRS has sufficient capacity to remove the decay heat. It is found that the natural circulation can be maintained under different pressures, the heat load sudden change condition as well, which indicates that the SPHRS natural circulation has good stability.

AP1000® SBLOCA simulations with TRACE code

The AP1000® is an advanced pressurized water reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The performing of such systems must be evaluated through the performance of experiments and simulations with a variety of thermal–hydraulic codes. This paper presents the results which has been obtained for different SBLOCA break sizes with the best estimate TRACE V5.0 patch 2 thermal–hydraulic code and their comparison with those obtained by Westinghouse with NOTRUMP code. The main results show that TRACE code predicts a similar trend in all sequences with some differences that are expected to be an issue of the more conservative models and hypothesis assumed in the SBLOCA licensing analysis performed with NOTRUMP. Some particular characteristics of this reactor are also shown in this paper such as the importance of core makeup tank (CMT) and passive residual heat removal (PRHR) system for core cooling in case of small break sizes and the behavior of the plant in case of availability of half of the total passive safety injection systems which is the case of the double-ended direct vessel injection line break (DEDVI).

Viability of uranium nitride fueled high-conversion PWR

Light water breeder reactors have been examined periodically for decades, and considerable research into the neutronics and thermal hydraulics of several designs has been performed. Currently, there is an interest in reduced moderation boiling water reactors in Japan and in the United States, where Department of Energy (DOE)-sponsored university research programs are examining proposed breeding BWRs.This study assesses the thermal hydraulic and neutronic feasibility of a nitride fueled pressurized water reactor (PWR) breeder design. Because of the higher fuel density, the use of nitride fuel would be preferable to the traditional oxide fuel for a high conversion PWR design. To illustrate the advantage, a design that uses large hexagonal assemblies was considered. The design has 14 inner seed pin rows and 4 outer blanket pin rows, leading to 505 inner seed pins and 366 outer blanket pins per assembly. Each assembly contains 6 thimble locations for control rods. The same assembly type can be used at all locations in the core. For this study a fuel pin diameter of 1.2 cm was used with a gap of 1 mm between pins. The seed regions had 12.75% percent of the heavy metal as fissile plutonium (Pu-239 and Pu-241), hosted in depleted uranium (0.25% U-235). The plutonium vector for the seed region was 2.7% Pu-238, 47.9% Pu-239, 30.3% Pu-240, 9.6% Pu-241, 8.5% Pu-242 and 1.0% Am-241. The Blanket regions contained only depleted uranium with 0.25% U-235.This assembly was modeled with the Monte Carlo depletion code Serpent to determine the initial to final fissile inventory ratio (FIR). The as-specified assembly model did not achieve an FIR value above 1.0, so the water in the thimble regions was changed to void, to lower the H/HM ratio. This led to FIR values above 1.0 for the oxide, the 85% theoretical density nitride (N85) and the 95% theoretical density nitride (N95). All designs had an FIR of 1.03 at 35 MWd/kgHM. However, the single batch discharge burnup of the assembly was much higher for the nitride cases. The burnups were 15.6, 24.5 and 32.2 MWd/kgHM for the oxide, N85 and N95 respectively.The KfK correlation for critical heat flux was applied to calculate the minimum departure from nucleate boiling ratio (MDNBR), whose value was set to be 1.45. With an acceptable core pressure drop of 400 kPa and acceptable vibration magnitude, the maximum power in an assembly and the maximum flow rate were determined. A nitride fuel had a larger MDNBR than the oxide fuel. This increase is likely a result of spectrum hardening, which was found to reduce the fractional amount of power generated in the seed region, where departure from nucleate boiling (DNB) occurs. Because of the longer neutron mean free path in the harder spectrum, more fission events occur in the blanket. The coolant density coefficient of the nitride fuel was found to have positive coefficient values in voided thimble tube cases. However, the power coefficient was negative and has the ability to limit power increases. The radially reflective assembly boundary may have made the analysis more conservative than full core analysis.

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.

Westinghouse AP1000 Plant, a Generation III+ Reactor

In a first part, this paper describes the AP1000® pressurized water reactor (PWR) design and how the AP1000 plant fundamental objectives of safety, simplification and standardization were applied throughout the design of the AP1000 plant.The second part of this paper focuses on specific safety and licensing topics: the AP1000 plant robustness for extreme events that may lead to catastrophic loss of infrastructure (such as the event that occurred at Fukushima Dai-ichi in March 2011) and the AP1000 plant compliance with the safety objectives for new plants released by the Western European Nuclear Regulators’ Association (WENRA).

Netronic Design of Small Long-Life PWR Using Thorium Cycle

A small long-life core loaded with thorium fuel and 231Pa as burnable poison material has been performed in Pressurized Water Reactor (PWR). Thorium cycle fuel has higher conversion ratio in the thermal spectrum domain and lower reactivity swing than the Uranium-Plutonium cycle fuel. 231Pa have very large capture cross section that can pressed reactivity in the beginning of life. The neutronic analysis result of infinite cell calculation shows that mixed nitride is better than oxide and carbide in thorium fuel system. In the present study we consider thorium nitride system with 3 ~ 8 % 233U percentage and 0.2~ 7% 231Pa as fuel for small PWR and can be burn up for the long time. The purpose of the study is to optimize the design of 350MWt PWR which can be operated without refueling in 10 years The core was designed by cylindrical two-dimension R-Z (radial and axial). The multigroup diffusion and Burn-up analysis was performed by SRAC-CITATION code using libraries based on JENDL 3.2. By using this concept, small PWR can be designed for long time operation with reduced excess reactivity until under 1 % and flatted power distribution during its operation.

Development of a Three-Dimensional Pseudo Pin-by-Pin Calculation Methodology in ANC

Advanced cores and fuel assembly designs have been developed to improve operational flexibility and economic performance and to further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000® plant of Westinghouse next-generation pressurized water reactor design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. The mixture loading of fuel assemblies with significant neutron spectra causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art three-dimensional (3-D) pin-by-pin calculation methodology (P3C) and successfully implemented it in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3-D P3C methodology along with its application and validation are discussed in the paper.

APR에서 사용하는 SDS 특성 분석에 관한 연구

안전감압시스템(SDS)의 주요 기능은 시스템의 압력이 운전 제한치를 초과하는 비정상적인 상황이 일어날 경우에 빠른 시간 내에 안정적으로 발전소의 압력을 감압시켜 발전소를 정상화하는데 그 목적이 있기 때문에 이러한 기능을 만족시키기 위한 안전감압시스템의 설계 인자가 무엇인가를 찾고자 하였다. 시스템의 압력을 조절하는 가압기 상단부분의 안전 감압밸브를 열어 증기를 배출함으로 시스템의 압력을 낮추는데 증기 배출을 위한 배관 설계에 있어서 배관내의 과도 유동 상태를 실제 발전소 여건에 맞도록 전산 프로그램을 개발하여 안전감압시스템의 배관에서의 증기의 과도 유동 해석을 수행하였고 그 결과를 토대로 안전감압시스템의 배관 및 배관 지지대 설계에 활용하고자 하였다. The purpose of this study analysis is evaluation of Safety Depressurization System function on PWR plant. The Safety Depressurization System is a system which releases steam from the pressurizer of Reactor Coolant System(RCS) and depressurizes RCS. There is a similar system in Boiling Water Reactor(BWR) type plant but the SDS with In-Containment Refueling Water Storage System(IRWST) is a new system to Pressurized Water Reactor(PWR). The current worldwide trend in a new PWR design is to implement a SDS with IRWST because of the advantages in installing IRWST. There have been also trials to install a SDS with IRWST to newly ordered domestic plant was that the domestic technical level for analysis and design of the system was not sufficient to do work.

Automatic multi-cycle reload design of pressurized water reactor using particle swarm optimization algorithm and local search

An automatic multi-cycle core reload design tool, which searches the fresh fuel assembly composition, is developed using particle swarm optimization and local search. The local search uses heuristic rules to change the current search result a little so that the result can be improved. The composition of the fresh fuel assemblies should provide the required cycle energy and satisfy the constraints, such as the hot zero power moderator temperature coefficient and the hot channel factor. Instead of designing loading pattern for each FA composition during search process, two fixed loading patterns are used to calculate the core status and the better fitness function value is used in the search process. The fitness function contains terms which reflect the design objectives such as cycle energy, constraints, and fuel cost. The results show that the developed tool can achieve the desire objective.

Optimization of Thermal-Hydraulic Reactor System for SMRs via Data Assimilation and Uncertainty Quantification

This paper discusses the utilization of an uncertainty quantification methodology for nuclear power plant thermal-hydraulic transient predictions, with a focus on small modular reactors characterized by the integral pressurized water reactor design, to determine the value of completing experiments in reducing uncertainty. To accomplish this via the improvement of the prediction of key system attributes, e.g., minimum departure from nucleate boiling ratio, a thermal-hydraulic simulator is used to complete data assimilation for input parameters to the simulator employing experimental data generated by the virtual reactor. The mathematical approach that is used to complete this analysis depends upon whether the system responses, i.e., sensor signals, and the system attributes are or are not linearly dependent upon the parameters. For a transient producing mildly nonlinear response sensitivities, a Bayesian-type approach was used to obtain the a posteriori distributions of the parameters assuming Gaussian distributions for the input parameters and responses. For a transient producing highly nonlinear response sensitivities, the Markov chain Monte Carlo method was utilized based upon Bayes’ theorem to estimate the a posteriori distributions of the parameters. To evaluate the value of completing experiments, an optimization problem was formulated and solved. The optimization addressed both the experiments to complete and the modifications to be made to the nuclear power plant made possible by using the increased margins resulting from data assimilation. The decision variables of the experiment optimization problem include the selection of sensor types and locations and experiment type imposing realistic constraints. The decision variables of the nuclear power plant modification optimization problem include various design specifications, e.g., power rating, steam generator size, and reactor coolant pump size, with the objective of minimizing cost as constrained by required margins to accommodate the uncertainty. Since the magnitude of the uncertainty is dependent upon the experiments via data assimilation, the nuclear power plant optimization problem is treated as a suboptimization problem within the experiment optimization problem. The experiment optimization problem objective is to maximize the net savings, defined as the savings in nuclear power plant cost due to the modified design specifications minus the cost of the experiments. Both the experiment and the nuclear power plant optimization problems were solved using the simulated annealing method.

Korean Development of Advanced Thermal-Hydraulic Codes for Water Reactors and HTGRs: Space and Gamma

The Korea nuclear industry has been developing the thermal-hydraulic system analysis Safety and Performance Analysis CodE (SPACE) and the GAs Multicomponent Mixture Analysis (GAMMA) code for safety analysis of pressurized water reactors (PWRs) and high-temperature gas-cooled reactors (HTGRs), respectively. SPACE will replace outdated vendor-supplied codes and will be used for the safety analysis of operating PWRs and for the design of an advanced PWR. SPACE consists of up-to-date physical models of two-phase flow dealing with multidimensional two-fluid, three-field flow. GAMMA consists of multidimensional governing equations consisting of the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of n species. GAMMA is based on a porous media model so that thermofluid and chemical reaction behaviors in a multicomponent mixture system and heat transfer within solid components, free and forced convection between a solid and a fluid, and radiative heat transfer between solid surfaces can be dealt with. GAMMA has a two-dimensional helium turbine model based on the throughflow calculation and a coupled neutronics-thermal-hydraulic model. Extensive code assessment has been performed for the verification and validation of SPACE and GAMMA.

Passive Cooldown Performance of Integral Pressurized Water Reactor

The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.

Comparative Study on the Safety Characteristics of Fission and Fusion Power Plants in Korea

The Korean fusion technology roadmap specifies the construction of a fusion power plant at demonstrative scale by 2030. Obviously, the safety requirements for demonstration fusion reactors will be quite different and more stringent than that of experimental reactors. Nevertheless, the regulatory framework for such reactors was not fully matured due to the limited resources and the lack of technical feasibility in Korea. Sharing with the motivation, this research investigated and compared the safety characteristics of fission and fusion power plants to facilitate designing of engineered safety features. Korea has gained a vast experience over the last 30 years, regarding design, construction and operation of both pressurized light and heavy water reactors, which is useful to address the attributes for fission power plants. In case of fusion reactor technology, the operational experiences with ITER and K-STAR can be referred, considering their demonstration scale. Comparative study was performed in top-down manner. We compared the top requirements such as safety principles and defense-in-depth for fusion and fission power plants. The inherent safety parameters such as the reactivity feedback coefficients of fission power plants were investigated how these parameters would be represented in fusion power plants. The limits for operating conditions for a fusion reactor were investigated to recognize important parameters which would contribute to nuclear safety or, more specifically accident prevention. For the accidents beyond the operation limits, the need of engineering safety features was found indispensable for accident mitigation. However, it is anticipated that the engineering safety features for fusion reactors will be reduced in number, size, type, and safety-margin because the total amount of hazardous material is much lower as compared to fission reactors. Finally we proposed the table of contents of safety analysis report for fusion power plants borrowing the basic structure from the safety reports on fission reactors. The outcome of this study helps to prioritize research projects to be devoted for analyzing the safety of demonstration fusion plant, and to develop design and regulatory framework in South Korea.