富集硼酸在压水堆一回路水化学中的应用研究

压水堆核电厂一般采用天然硼来控制反应性。在核电厂实施长循环燃料管理后,寿期初硼浓度较高,增加了水化学控制的压力。本文开展了富集硼酸(EBA)在一回路水化学中的应用可行性及其对相关水质处理系统的影响分析。研究表明一回路采用EBA有助于降低结构材料的腐蚀和堆外辐射场,提高在役核电厂的经济性。     

富集硼酸在核电厂一回路冷却剂中的应用研究

随着压水堆核电厂逐渐向长周期燃料循环转变,堆芯功率密度不断提高,燃耗不断加深,一回路冷却剂水化学控制也变得更为复杂和困难。对核电厂一回路富集硼水化学进行计算分析,结果表明富集硼酸的使用,可降低冷却剂硼浓度,提高p H值;10B富集度在40%以上的富集硼酸能维持堆芯运行于p H值7.2~7.4。     

压水堆活化腐蚀产物源项~(58)Co和~(60)Co敏感性研究

以某压水堆核电厂为例,采用CORA程序分析压水堆核电厂一回路材料组成、蒸汽发生器传热管材料钴含量、冷却剂氢氧化锂浓度、净化效率和反应堆运行功率等因素变化对一回路腐蚀产物58Co和60Co活度浓度的影响。计算结果表明:通过限制蒸汽发生器传热管材料中钴元素的含量、提高冷却剂中氢氧化锂浓度、提高冷却剂净化效率和降低功率等措施可以有效降低活化腐蚀产物的活度浓度,为压水堆核电厂辐射剂量控制提供参考。     

压水堆核电厂一回路冷却剂系统中锕系核素质量评估方法研究

为评估压水堆核电厂燃料包壳破损时的工作人员辐射风险和燃料包壳破损程度,基于特征物理量建立一回路冷却剂系统中锕系核素质量评估方法。本文基于锕系核素的生成和迁移机理,建立了一回路冷却剂系统中锕系核素的平衡方程组,并选取3种易监测的特征物理量用以评估锕系核素向一回路冷却剂系统的释放量及其分布,并建立了一回路冷却剂系统中锕系核素质量的评估方法。然后分别采用国内在役压水堆核电厂无燃料包壳破损和有燃料包壳破损的实测数据对建立的评估方法进行了验证,验证结果表明：建立的评估方法可在无燃料包壳破损和有燃料包壳破损的情况下对一回路冷却剂系统中锕系核素质量进行评估,评估结果和预期符合。本文研究成果可为压水堆核电厂运行期间一回路冷却剂系统中锕系核素质量及其分布评估提供指导,从而优化后端的工作人员防护措施,降低辐射风险。     

标准导读

正NB/T20142-2012《压水堆核电厂一回路系统及设备化学去污》。本标准规定了压水堆核电厂一回路系统及设备化学去污的去污剂、去污方法、去污实施、效果评估等方面的技术要求。本标准适用于压水堆核电厂一回路系统及设备表面在线或离线化学去污,其他系统或设备表面的放射性污染去除,亦可参照执行。定价:30.00元NB/T 20001—2013《压水堆核电厂核岛机械设备制造规范》。本标准给出了GB/T16702规定范围内的压水堆核电厂核岛机械设备制造过程中的标识、切割和不作焊补的修     

压水堆核电厂一回路冷却剂泄漏率计算的优化

一回路冷却剂的泄漏率是压水堆核电厂放射性控制相关的一个重要物理量,需要定期进行监测.但由于目前国内核电厂对其研究较少,其测量和计算中存在一些不足.本文立足于现场运行实际,通过对秦山第二核电厂一回路泄漏率的分析计算,总结和完善了压水堆核电厂一回路冷却剂泄漏率的计算方法.     

田湾核电厂一回路水化学优化与辐射源项控制

压水堆核电厂一回路辐射场的影响因素众多,主要有两个方面,一是系统设备的材料,二是一回路水化学参数。本文对田湾核电厂溶解H₂、硼碱协调曲线控制、注Zn技术、停堆氧化对辐射源项的影响进行探讨,最后提出建议。     

压水堆核电厂一回路抽真空排气方法和装置研究

介绍了压水堆核电厂反应堆一回路抽真空排气方法,以及由带密封环反应堆压力容器临时顶盖、抽真空排气台架组成的抽真空排气装置设计方案和应用过程。利用该装置,在国内首次实现核电厂大修低低水位期间的反应堆一回路抽真空排气,取消了原有动排气过程,可缩短大修关键路径时间10余小时,降低反应堆冷却剂系统主泵损坏风险,提高电厂运行经济性和安全性。     

压水堆核电厂核岛硼结晶问题分析及处理对策

综合介绍了压水堆核电厂核岛硼结晶问题的概况,阐述了硼结晶产生的机理及腐蚀危害,结合实践经验及国内外调研,制定了不同类别硼结晶问题的处理标准及对策,并成功在实践中应用,同时提出了压水堆核电厂在减少硼结晶方面的后续改进方向。     

硼表硼浓度算法设计与仿真验证

为解决压水堆核电厂离线式硼表（OFBM）测量的核岛一回路冷却剂总硼浓度相对化学滴定硼浓度的偏差超标问题,对OFBM的硼浓度算法进行了分析,并对硼浓度偏差的影响因素和产生的原因进行了分析。分析结果表明:硼浓度偏差的主要影响因素是核岛一回路冷却剂中的10B浓度和10B丰度;OFBM硼浓度算法忽略了核岛一回路冷却剂的10B丰度变化是造成硼浓度偏差的主要原因。同时设计了一种能够跟踪核岛一回路冷却剂10B丰度的新硼浓度算法,新算法实现了10B浓度计算功能。最后建立了新硼浓度算法的仿真模型,并基于2台M310核电机组OFBM的标定数据对新硼浓度算法的计算准确度进行了验证,仿真结果表明,新硼浓度算法的计算准确度满足OFBM设计规格要求。      

Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD codes CFX-5 and Trio_U

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the downcomer of a pressurized water reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields.An experiment with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water was selected for validation of the CFD software packages CFX-5 and Trio_U. Two similar meshes with approximately 2 million control volumes were used for the calculations. The effects of turbulence on the mean flow were modeled with a Reynolds stress turbulence model in CFX-5 and a LES approach in Trio_U. CFX-5 is a commercial code package offered from ANSYS Inc. and Trio_U is a CFD tool which is developed by the CEA-Grenoble, France.The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: at higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this propagation. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. Both CFD codes were able to predict well the observed flow patterns and mixing phenomena.     

The inherent boron dilution mechanism in pressurized water reactors

The purpose of this paper is to describe a mechanism that inherently causes boron dilution in pressurized water reactors (PWRs). The phenomenon is due to the fact that boric acid does not markedly dissolve into steam. This is relevant for transient and accident situations in PWRs where decay heat removal is accomplished by coolant vapourization and condensation, which inherently leads to formation of dilute plugs in the primary. In particular, it is found that inherent dilution will be inevitable for a range of small break loss of coolant accidents (SB LOCAs), with maximum amount of total diluted coolant mass exceeding 20 tonnes for a modern 1300 MWe PWR equipped with U-tube steam generators. A simple analysis of dilute plug motion during the late phases of a SB LOCA and core response to boron dilution shows that the damaging potential might extend to widespread fuel failures. Other transients and accidents are also discussed from the point of view of inherent dilution. Some possible remedies to the problem, as well as suggestions for further research, are presented.     

压水堆核电站一回路活化腐蚀产物源项控制措施探讨

材料替代和一回路水化学控制是降低活化腐蚀产物源项的主要措施。本文介绍M310、AP1000和EPR三种压水堆核电站一回路水化学优化情况,比较三种压水堆一回路活化腐蚀产物源项,分析探讨水化学优化对源项降低的影响,最后对国内压水堆核电站一回路水化学优化提出建议。     

Modeling the activity of I and Cs in the primary coolant and CVCS resin of an operating PWR

Mathematical models have been developed to describe the activities of ¹²⁹I and ¹³⁷Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of ¹³⁷Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries.     

Boron transient effects on the behavior of IRIS reactor using dynamic modeling

Small Modular Reactors (SMR) are considered as having several advantages over typical nuclear reactors under various specific conditions. They are thought to be installed in countries with small or medium power grid, in which a large power plant is not necessary or in isolated communities far from distribution centers. A plenty of developing countries are in this situation, so that a significant demand on this type of reactor is expected in a near future. The IRIS reactor is the top-front of SMRs, making its complete development very attractive, since it can fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is an integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes when compared with a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In light water reactors, a solution of boric acid is used in the coolant of the primary loop to absorb neutrons, aiming to adjust the reactivity of the reactor. A significant decrease in the boron concentration in the core might lead to a considerable power excursion. Several studies on PWR have established correlations between power excursions and deficiencies in homogenization of boric acid diluted in the coolant. The IRIS reactor, due to its integral configuration, does not possess a spray system for boron homogenization which may cause power transients. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics model for power generation. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were inserted into the SIMULINK and the code was validated by comparing with RELAP simulations for a transient of feedwater reduction in the steam generators. Furthermore, the current paper looks for studying and developing a dynamic model for calculating the variations in the boric acid concentration. Then, a simplified model for boron dispersion was implemented into the code MODIRIS to simulate power transients which occur due to variations in the boron concentration in the primary loop of the IRIS reactor. The results for boron concentration, inserted reactivity and steam production showed a good precision and represented the expected behavior very well in the range of operational transients.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

Ex-vessel transport and mixing of a deborated slug in a PWR primary geometry

The transport and mixing of a slug of deborated water in a lowered loop PWR is modeled by partitioning the volumes of the primary system according to chemical rector theory. Piping is modeled as plug flow volumes while the steam generator outlet plenum and the reactor coolant pumps are modeled as backmixed volumes. This simple approach provides a good representation of the transport and mixing phenomena outside the reactor vessel. The proposed methodology can be used to generate initial and boundary conditions for separate effects tests and CFD computations for the reactor vessel complex geometry. The decoupling of the ex-vessel primary system greatly enhances the resolution of boron dilution transient issue.     

Design of advanced integral-type marine reactor, MRX

The basic concept of an innovative advanced marine reactor with a passive safety system, MRX (Marine Reactor X) has been established for the primary application to ship propulsion. The design goals of the reactor system, to be lightweight and compact, and to be enhanced in safety and reliability, are achieved with adoption of new technologies. The MRX is of the integral-type PWR with 100 MW of thermal output. Adoption of a water-filled containment makes the MRX extremely lightweight and compact. The total weight and volume of MRX is about 1600 tons and 1210 m³, which is half that of the first Japanese nuclear ship, ‘Mutsu’, reactor, although the reactor power of MRX is three times greater than that of the ‘Mutsu’ reactor. Numbers of active components in the reactor system are greatly reduced, compared with loop type PWRs, by adopting an integral type reactor and the passive system. Safety was evaluated by both experiments and analyses. Core damage occurrence frequency estimated by probability safety analysis (PSA) is of two orders smaller than those of existing PWRs. Feasibility study on economics is conducted by comparing the total operation costs of a nuclear container ship installing the MRXs with a diesel engine ship. The nuclear ship has the advantage for greater speed and larger amounts of cargo carried.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

一种铅-铋合金冷却快堆的高放废物积累量研究

目前商用压水堆积累了大量的长寿命高放废物,放射毒性强,衰变时间漫长,对环境和人类构成了长期威胁,作为6种第四代核能系统堆型中的一种,铅基冷却快堆在减少长寿命高放废物产生方面具有优势。基于此本文提出了一种热功率为300 MW的铅-铋合金冷却快堆设计。利用MCNP程序对反应堆堆芯进行建模并计算了堆芯在寿期初的主要物理参数,详细分析了燃耗过程中长寿命高放核素的积累量,并与一般压水堆长寿命高放核素的积累量进行了比较。结果表明,对主要关心的次锕系核素,铅-铋合金冷却快堆的产生量远小于压水堆的,而长寿命裂变产物的产生量与压水堆的相当。总体来说,铅-铋合金冷却快堆产生的长寿命高放废物总量小于压水堆的,可看出铅-铋合金冷却快堆在减少长寿命高放废物产生方面更具有竞争性。