Development of the package-reactor (2) -core characteristics-

A new small reactor concept named the Package-Reactor has been jointly developed by Hitachi, Ltd. and Mitsubishi Heavy Industries, Ltd. The reactor technology was built from that of conventional LWRs. The reactor core consists of 12 cassettes containing fuel rods with a similar design to that of PWR fuel rods. Each cassette has about a 0.4 m outer diameter and they are fixed with about 0.5 m pitch to each other in the atmospheric pressure condition. A tube-type control cluster was developed. It can decrease the rise of reactivity for the one-rod-stuck condition. An advanced cassette design was studied in which the down-comer is placed at the center of the fuel region. This concept, which improves neutron economics and the cold shutdown margin, will increase the marketability of the Package-Reactor. An operation period of more than 8 years can be achieved with UO₂ fuel enrichment of 5.0wt%.     

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

1 Introduction With respect to the inherent safety of nuclear re- actors, application of passive systems/components including natural circulation phenomena as the main mechanism is intended to simplify the safety-related systems and to improve their reliability, to reduce the effect of human errors and equipment failures, and to provide more time to enable the operators to prevent or mitigate serious accidents. Natural circulation is the main mode of heat removal for removing decay heat from t…     

HPLWR equilibrium core design with the KARATE code system

The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of neutronic core design within the framework of the "HPLWR Phase 2" FP-6 and the Hungarian “NUKENERG” projects. As the coolant density along the axial direction shows remarkable change, coupled neutronic-thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the assemblies. A parametrized diffusion cross section library was prepared for the HPLWR assembly with the MULTICELL neutronic transport code. The parametrized cross sections are used by the KARATE program system, which was verified by comparative Monte Carlo calculations. Preliminary loadings of the HPLWR core were assessed, which contain insulated assemblies with Gd burnable absorbers. The fuel assemblies have radial and axial enrichment zoning to reduce hot spots.     

300MW压水堆核电厂堆芯反应性控制组件的设计和研究

总结了我国３００ＭＷ压水堆核电厂堆芯反应性控制组件设计的基本经验。针对控制棒的主要失效模式，讨论了关键的技术问题，对于首次使用的的硼硅酸盐玻璃可燃毒物，着重研究了抗强辐照性能，以于次级和初级中子源棒，分别阐述了重要的内压问题和有关的安全性能要求.?     

Three pass core design proposal for a high performance light water reactor

The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280 °C at the reactor inlet to 500 °C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step is foreseen in a central “evaporator” and two further steps in a first and a second superheater surrounding it. The coolant flow scheme includes upward and downward flow through the core with intermediate mixing in chambers above and below the core to eliminate hot streaks. A preliminary single channel analysis, concentrating on an average flow channel and on the hottest one only, indicates that such core design can match the limits of cladding materials available today. Even though the resultant pressure drop of the coolant will be higher than usual, it is expected that the assembly boxes can be designed with acceptable deformations.     

Estimation of sensitivity coefficients of core characteristics based on reduced-order modeling using sensitivity matrix of assembly characteristics

We propose an estimation method of sensitivity coefficients of core neutronics parameters based on a multi-level reduced-order modeling approach. The idea is to use lower-level models to identify the dominant input parameter variations, constrained to the so-called active subspace, which are employed to determine the sensitivity coefficients of the core neutronic parameters. In our implementation, the lower-level model is represented by two-dimensional assembly calculations, which are employed in the preparation of the few-group cross-sections for core-wide calculations. The active subspace basis is estimated using the singular value decomposition of sensitivity matrices of assembly neutronics parameters. In numerical verification calculation, sensitivity coefficients of core characteristics for a typical three-loop PWR equilibrium-cycle are estimated using the proposed method and the direct method. Comparison of these two results shows that the proposed method well reproduces the results obtained by the direct method with lower calculation costs. Through the verification calculations, applicability of the proposed method to practical light water reactor analysis is confirmed.     

Development of advanced core noise monitoring system for BWRS

A BWR core noise monitoring system is developed for addressing core anomaly problems in future advanced core operation. In order to monitor in-core status from a limited number of signals, various up-to-date signal processing algorithms are introduced to compensate for a lack of information. These algorithms, such as independent component analysis, factor analysis and model based parameter estimation, are demonstrated to be effective through real plant data analysis to evaluate core and regional stability index, reactivity coefficients and core flow rate. Through these practices, we demonstrate that the core noise monitoring system is an effective general platform for providing a variety of monitoring tools to meet the requirements in future advanced core operation.     

中国先进研究堆稳态自然循环能力分析

针对中国先进研究堆(CARR)的结构和运行特点,开发了CARR自然循环能力计算程序,计算得到了不同池水温度条件下CARR自然循环能力,并分析了池水温度对CARR自然循环特性的影响:自然循环冷却剂流量随池水温度的升高而增大,但自然循环能力(带走的堆芯功率)随池水温度升高而降低.基于理论推导和程序计算结果,提出了一个适用于预测不同池水温度下CARR自然循环流量和堆芯功率的简单关系式,该关系式预测值与程序计算结果误差小于±10%.     

Development of medium and small sized reactors: DMS

Nuclear power is expected to become the main source for electric power generation in Japan for the reasons of energy security and prevention of CO₂ emission. In addition, the slowdown of recent electric power demand and the liberalization of the electric power market are accelerating medium and small sized reactor development. (Hida and Ito, 2003) Furthermore, the needs of medium and small sized reactors have become greater in foreign countries where electric grid systems are weak. Under these circumstances, Hitachi has developed DMS's (Double MS: Modular Simplified & Medium Small Reactors) as 400 MWe class LWR's supported by The Japan Atomic Power Company. (Moriya et al., 2003) In addition, DMS's have been designed based on proven technology that requires no large-scale development, and can therefore be introduced in the market in near future.     

中国先进研究堆停堆冷却措施及其主要特点

停堆后冷却问题是中国先进研究堆(CARR)重要的安全问题之一.冷却措施的实施对CARR的安全和建设投资有较重要的影响.CARR采用停堆初期的强迫循环及停堆后期全堆芯自然循环相结合的策略实现正常停堆和事故停堆后的堆芯冷却.停堆冷却的过程具体分为主泵大质量惯性飞轮惰转强迫冷却、应急堆芯冷却系统强迫冷却、自然循环功能部件动作实现全堆芯自然循环3个阶段.3个阶段既相互衔接又相互独立,每个阶段各有特点.停堆冷却策略的实施证明,CARR停堆冷却过程是可靠、有效、合理的,符合先进研究堆的发展趋势.     

先进轻水堆核电站的用户技术设计要求

自美国电力研究所用 求文件和欧洲用户要求文件发表以来,目前已被用于好几个先进轻水堆核电站的设计,有关国家的核安全管理当局也对这些文件持肯定的态度。本文重点描述了ＵＲＤ和ＥＵＲ的文件结构,所阐述的有关安全 及所建立的主要定量安全要求,并简便介绍了有关核安全管理当局对这些用户要求文件的看法。     

轻水堆严重事故及可能的缓解措施

现有概率安全评价指出，常规轻水堆的堆芯熔化频率及安全壳失效，放射性大量释放的频率都是是很低的。但这些风险对于下一代先进轻水堆说是不能忽力听，近年来西方对下一代先进轻水堆的安全目标作了更高的要求，即在严重事故的条件下，仍然能保证安全壳的完整性，而无需采取应急措施，这就要求对严重事故现象可有足够的认识，以便对严重事故设置相应的缓解措施，本文简述了严重事故的物理现象，机理及可能的缓解策略，综述了这方面的     

Investigation of burnable poisons effects in reactor core design

Burnable absorber rods (BAR) and chemical shim are the main control poisons that are used in the core for improving the reactor behavior and satisfying the safety criteria during the core life time. These poisons have several constraints, criteria, advantages and also disadvantages from the safety and operation points of view; and these characteristics depend on the concentration and distribution of mentioned poisons in the reactor core. Therefore, understanding their effects on the reactor core behavior, especially the mutual interaction between them, is a crucial issue in reactor core design procedure. In this study, the influences of the burnable poisons on the main parameters of the reactor such as multiplication factor, burnup, soluble poison concentration, moderator temperature coefficient and power peaking factor over the reactor life time are investigated. The VVER-1000 reactor was selected for this investigation.     

Benchmarking verification of the control rod effects on the MTR core parameters using the MTR-PC and MCNP codes throughout 3D core modeling and rod-drop experiment

Verification of nuclear data and codes is highly recommended for reactor safety analyses. In this research, MTR-PC package and MCNP 4C are used as deterministic and Monte Carlo simulation code, respectively. They are taken into account safety parameters as a function of control rods. Control absorbers classified in two different types. The first called Shim Safety Rods (SSR) made of an Ag–In–Cd composition, and the second one is the Fine Regulating Rod (FRR) made of Stainless Steel. One startup test of a 5 MW Material Testing Reactor (MTR) is simulated throughout the 3-D core modeling. It checks the overall simulating performance. Reactivity worth effects of control rods are calculated and benchmarked against rod-drop experimental results. Both types of codes are taken into account the integral and differential behavior of reactivity worth effects. The startup state has been simulated very carefully, and also total reactivity worth effect of control rods differs from the measured value less than 1.1% (in pcm). Results are in good agreement and highly reliable to calculate the total reactivity worth effect of control rods, and to simulate the startup state of MTRs as Best Estimate (BE) tools.     

先进压水堆非能动余热排出技术试验研究

总结了中国核动力研究设计院空泡物理和自然循环国家级重点实验室10年来开展的先进压水反应堆非能动余热排出技术试验研究和专用程序开发研究，提出了下一步开展相关工程研究的建议。     

脉冲堆热工水力设计程序研制开发

介绍了中国核动力研究设计院自主开发的脉冲堆热工水力设计程序系统。它包括脉冲堆自然循环分析程序（MC－FLOW）、堆芯热工水力分析程序（MC－THAS）和脉冲堆瞬态分析程序（MC－TRAN）。采用原型堆的数据对程序进行验证，其结果表明：脉冲堆热工水力设计程序系统满足热工水力设计的要求，能够可靠地用于西安脉冲堆的设计。     

压力管式反应堆非能动余热排出系统方案研究

本文针对压力管式钍基先进核能系统(TANES)提出了一种非能动余热排出(PRHR)系统方案。该方案利用两个回路的自然循环，将事故工况下的堆芯余热排出到最终热阱。利用RETRAN02程序，以全厂断电事故为设计基准事故，对TANES非能动余热排出系统的余热排出能力进行了计算。计算表明，TANES的PRHR系统能够将余热导向最终热阱并且保持冷却剂回路和慢化剂回路的压力低于设计限值。另外对诸如设备间高度差等因素进行了敏感性分析。     

Progress in the studies of the advanced plutonium fuel assembly

A new assembly concept, designated APA (for dvanced lutonium fuel ssembly), should make it possible to multi-recycle plutonium in pressurized water reactors. The basic idea is founded on the manufacture of a large plutonium thin annular fuel rod with an inert support, cooled on both faces. The absence of plutonium generation, combined with moderate fuel temperature should make it possible to achieve substantial burn-up fractions in these rods. The assembly is compatible with the internals of a Pressurized Water Reactor (PWR), and provides for permanent reversibility. Neutronic studies showed a compliance with actual safety/control criteria. A multi-recycling scenario was simulated for 84 years' operation with a 65 GW electrical power installed capacity, comprising forty-five 1450 MW electrical power PWRs, 32 of which are loaded with UO₂ and 13 with APAs. It showed that the plutonium inventory is controlled. Thermal-hydraulic studies showed one can find an annular rod geometry allowing one to respect both margins to Critical Heat Flux (CHF) during normal and accidental operations and void fraction limitations.     

一体化小型氟盐冷却高温堆瞬态特性分析

根据下一代核能系统的发展目标,提出了采用自然循环的一体化小型氟盐冷却高温堆的概念。利用修改后的RELAR5-MS系统分析程序,建立了一体化小型氟盐冷却高温堆模型,并得到其稳态特性参数。在此基础上,对其在满功率运行状态下的反应性引入事故和失热阱事故进行了分析。分析计算表明,在反应性事故工况下,由于自然循环的存在,堆芯冷却剂流量随着堆芯温度发生动态变化,最终达到新的稳态,燃料棒和冷却剂温度均处于安全限值范围内。在失热阱事故下,反应堆负反馈的特性使得堆芯功率逐渐降低并实现自动停堆,即使不考虑余热排出系统的作用,燃料组件和冷却剂温度上升缓慢,在140 h内,燃料棒和冷却剂温度均处于全限值范围内。结果表明,一回路采用自然循环冷却的一体化小型氟盐冷却高温堆具有良好的固有安全性。     

一体化小型自然循环反应堆流动非对称性机理分析

为研究一体化小型自然循环反应堆稳态条件下的流动非对称性机理,通过总结当前国际自然循环反应堆的主要设计特征,建立了自然循环典型三维分析模型,采用计算流体动力学（CFD）方法对竖直稳态下自然循环进行了数值模拟,并针对不同混流截面的速度和温度分布特性进行了分析。结果表明：较大尺度的自然循环系统在理想竖直稳态自然循环条件下,存在周向的流量和温度分布不均匀性,其中流量分配的不均匀性较温度分布情况更突出。竖直稳态自然循环的主流流量和温度均呈现偏心趋势,属于系统层面的偏差,而不只是局部流动分配不均衡问题。