Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

A study on reactivity insertion controlled LMR cores with metallic fuel

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity.

A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core.

The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.     

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of the program for the industry in Global Nuclear Energy Partnership initiatives, including a core in the first plant for demonstration and cores with enhanced TRU burning capability for the future plants. Both concepts for the first plant; low core height and large volume fraction of structure are deployed, seeking small TRU conversion ratio (CR)^* and small void reactivity which are crucial in the design, but different approaches. In this paper, the TRU CR and the sodium void reactivity have been approximated with a single equation in these concepts, based on the theoretical formula related to the chain reaction in the reactor and the calculation results. Shortening the core height and increasing the structure volume fraction will enhance TRU enrichment through increased neutron leakage and capture, which will reduce a ratio of U-238 to sum of Pu-239 and Pu-241 so that TRU CR decreases from 0.78 to 0.57. A small ratio of sodium loss to plutonium fissile will be effective also in the reduction of positive reactivity effect by spectral hardening. On the other hand, when this ratio and geometrical buckling of flux reduce, negative contribution by the neutron leakage becomes small. Theses relations related to the positive void reactivity have been formularized by the approximation with few parameters within several percent respectively as well as the TRU CR, indicating that one of dominating parameters is the ratio of sodium loss to plutonium fissile in the void reactivity at large fast reactors. * = (1 − net loss of TRU/loss of heavy metal).     

大型MOX燃料快堆钠空泡反应性微扰理论研究

钠空泡反应性效应是钠冷快堆核设计和安全分析的重要内容。本文基于多群节块扩散法,采用微扰理论推导出钠空泡反应性的计算方法,对1 000 MWe钠冷快堆MOX燃料堆芯的总钠空泡反应性、空间分布、物理分项进行了计算。结果表明,钠空泡反应性主要来源于中子泄漏的增加和能谱的硬化,两者一正一负,且空间分布规律相反,导致钠空泡反应性具有强烈的空间依赖性;对于所计算的MOX燃料堆芯钠空泡反应性高达3 $左右。计算和分析结果阐明了钠空泡反应性的产生机理和分布规律,可为低钠空泡的设计提供参考。     

重水冷却钍基长寿命模块化小堆概念设计研究

基于传统压水堆(PWR)技术,提出一种重水冷却的钍基长寿命模块化小堆(RMSMR)的概念设计方案,采用二维模型系统分析并对比了PWR和RMSMR的燃料类型、慢化剂类型等参数,获得反应堆各项中子学参数的变化机理;然后基于二维计算结果提出了最终的三维堆芯设计方案,并开展了初步的中子物理和热工安全分析。研究表明,RMSMR在设计上采用三区燃料布置来展平功率,采用钍-铀燃料维持了负空泡系数,通过布置增殖包层提高了堆芯的转换比(CR);RMSMR采用了重水冷却剂可以使中子能谱硬化,从而提高CR,减小寿期反应性波动,增加堆芯寿期;RMSMR能够在100 MW电功率下维持6 a的安全运行。本文研究可为新型反应堆的设计发展提供借鉴。     

Nuclear and thermal–hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.     

Three-dimensional core analysis on a super fast reactor with negative local void reactivity

Keeping negative void reactivity throughout the cycle life is one of the most important requirements for the design of a supercritical water-cooled fast reactor (super fast reactor). Previous conceptual design has negative overall void reactivity. But the local void reactivity, which is defined as the reactivity change when the coolant of one fuel assembly disappears, also needs to be kept negative throughout the cycle life because the super fast reactor is designed with closed fuel assemblies. The mechanism of the local void reactivity is theoretically analyzed from the neutrons balance point of view. Three-dimensional neutronics/thermal-hydraulic coupling calculation is employed to analyze the characteristics of the super fast reactor including the local void reactivity. Some configurations of the core are optimized to decrease the local void reactivity. A reference core is successfully designed with keeping both overall and local void reactivity negative. The maximum local void reactivity is less than −30 pcm.     

Neutronic feasibility design of a small long-life HTR

Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U-Battery. Increasing neutron moderation by replacing fuel blocks with graphite blocks and dispersing the graphite blocks in the reactor core are two effective ways to increase the fuel burnup and lifetime of the U-Battery. Water or steam ingress may induce positive reactivity ranging from 0 to 0.16 Δk/k, which further demonstrates that the U-Battery is under-moderated.     

小型长寿命核能系统燃料物理性能的研究

本文在简要说明世界上小型长寿命核能系统研究现状的基础上,提出了使用钍-铀燃料和铅-铋冷却剂构造小型长寿命堆芯的设想,并为此进行了一系列燃料物理性能的研究.对于长寿命核能系统的堆芯物理设计,使反应性随燃耗变动最小非常重要,同时应该尽可能地提高堆芯的燃耗以满足长寿命运行的需求.本文使用MCNP和MCBurn程序详细计算分析了使用不同的初始驱动燃料、不同栅格、燃料成分和类型、富集度条件下,燃料栅元的燃耗反应性变化等性能,并对其进行了能谱、转换比、富集度变化等方面的分析,经过对比初步确定了使用钍-铀燃料构造长寿命堆芯的物理条件,并以此为起点构造出一个堆芯,计算给出了反应性空泡系数等安全参数.     

Study on super-long-life cores loaded with minor actinide fuel

Super-long-life fast breeder reactor cores (SLLC) loaded with minor actinide (MA) fuel were designed aiming at continuous operation without refueling during plant lifetime and efficient reduction of MA nuclides (Np, Am and Cm). The feasibility was studied from nuclear and thermal characteristics. As a result, 1000 MWe and 300 MWe SLLCs with small reactivity change and power swing during plant lifetime were found to be feasible. MAs can be confined and transmuted in the reactor during plant life. A 1000 MWe SLLC can transmute MAs of 10 ton which come from 13 light water reactors (1000 MWe).     

Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B₄C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.     

Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Fore Configurations

Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ?2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25–40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient.     

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.     

Comparison and sensitivity analysis of the core characteristics of a sodium-cooled fast breeder reactor with 750 MWe output evaluated by JENDL-4.0 and ADJ2000R

Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the μ-average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross-sections of Am-241.     

Activities for 4S USNRC licensing

4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days.4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling.Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper.     

加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

运用MCNP与ORIGEN2耦合计算程序COUPLE，对加速器驱动的次临界系统（ADS）钠冷金属燃料快堆堆芯进行稳态与燃耗计算，比较分析次锕系核素（MA）非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明，对比均匀布置，非均匀布置具有更高的MA嬗变率与嬗变支持比，在反应性参数方面导致多普勒效应与有效缓发中子分额降低，钠空泡效应增大，在堆芯功率分布与加速器束流功率方面没有明显变化。     

Thermohydraulic transient studies of the Chinese 200 MWe HTR-PM for loss of forced cooling accidents

Pressurized and Depressurized Loss Of Forced Cooling (PLOFC and DLOFC) are two important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behaviors during LOFC can provide useful reference to the physics, thermohydraulic and structure designs of the reactor core, and can also verify the design of the Residual Heat Removal System (RHRS). The 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China, is characterized by its inherent safety features, such as shutdown ability via negative temperature coefficients of reactivity, passive mechanism of decay heat removal and so on.     

Use of Thorium-Plutonium MOX in the inner pins of CANDU fuel bundles

This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters.The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO₂ fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO₂. This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case.     

新型快堆—铅冷快堆的堆物理特征

讨论了以重金属铅作为冷却剂的快中子核反应堆的基本堆物理特征。铅由于其慢化截面小，输运截面低，对于快中堆带来一系列的特点，铅冷快堆具有较大的冷却剂－燃料体积比，较小功率密度，较硬的中子能谱，负的空泡反应性效应和负的温度反应性系数。这些特性提示了铅冷快堆具有固有安全性的多种因素。     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.