船用堆大破口失水事故放射性后果分析

本文以严重事故分析程序MELCOR为计算工具,建立了某型船用堆的计算模型,研究了某型船用堆发生冷段双端断裂大破口失水事故的源项行为及放射性后果。分析了惰性气体Xe与挥发性气体CsI的释放、迁移和舱室分布规律,并对通风系统投入时机进行研究。结果表明:为保证堆舱临舱的剂量辐射在剂量限值内,应于事故发生后10min内投入全船通风。否则,应于全身剂量和甲状腺剂量达到剂量限值前及时采取防护措施。     

International cooperation on control for safe operation of nuclear research reactors

The purpose of this paper is to present some of the main reasons, objectives, planning, and recent advances of an SCK·CEN/ININ joint project, which deals with the design and application of modern/expert control and real-time simulation techniques for the safe operation of a TRIGA Mark III research nuclear reactor. This project has been proposed as the first of its kind under a general cooperation agreement between the Belgian Nuclear Research Centre (SCK·CEN) and the National Nuclear Research Institute (ININ) of Mexico.     

中国高温气冷堆核电示范工程环境辐射影响初步分析

对我国高温气冷堆核电示范工程(HTR-PM)进行了环境辐射影响分析和评价.内容包括堆芯放射性总量的计算、正常运行工况下放射性核素的年释放量、事故源项的分析计算以及正常运行和事故情况下辐射剂量的估计.分析结果表明:正常运行工况下,HTR-PM放射性释放对公众成员可能产生的辐射剂量远低于我国目前的法规要求;设计基准事故情况下对公众成员可能产生的辐射剂量明显低于需要在场外采取隐蔽措施的通用干预水平.     

基于部队核事故后果分析的系统开发

通过分析国内核设施事故后果评价技术现状及相关系统的功能特征,研究部队核事故后果评价技术的发展情况,比较两者在评价技术方法和功能模式上的异同点。在此基础上,针对前期研制的一种适用于部队核事故后果评价的软件系统,在功能组成、开发应用和优化升级等方面进行了探讨,为部队核武器事故后果评价技术发展提供了启示。     

核事故应急放射性后果评价模式的有效性

放射性后果评价模式的验证和确认是目前开发评价模式中亟待解决的关键问题,本文介绍了模式验证和确认的实用方法,并针对模式验证和确认中的难点提出几点建议。     

Design,construction and simulation of a multipurpose system for precision movement of control rods in nuclear reactors

This article presents the design and implementation of a microcontroller-based system for the automatic movement of control rods in nuclear reactors of either power or research types. This system is controlled automatically, is linked to a personal computer system, and has manual controlling ability as well. The important features of this system are: automatic scram of the control rods, activation of alarm in emergency situations, and the ability to tune the control rod movement course both upwards and downwards. In this system, a small tank has been improvised as a coolant reservoir for pool type reactors such as Tehran Research Reactor and its water level is continuously adjusted by special sensors. Also, this system can be applied for controlling various types of control rods such as the regulating rods, safety rods and shim rods; can be connected to all reactor measurement tools and systems such as the period meter, power meter and flux meter; and can receive feedback signals from them. The devised system can be calibrated with these measurement tools by two special potentiometers in the related electronic board. The processes of this system have been simulated by the SIMULINK tool kit of MATLAB software and all responses of the system, including oscillation and transient responses, have been analyzed.     

Design and safety optimization of ship-based small nuclear power reactors

Design and safety optimization of ship-based nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X–Y–Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect and quasi-static approach is also employed to treat neutronic aspect during safety analysis.

The reactors are loop type lead–bismuth-cooled fast reactors with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to water–steam loop through steam generators. The power level is 100–200 MW th and excess reactivity is about 1$. Three types of core were investigated in the optimization process: balance, tall, and pancake with five values of Z–Y size ratio.

As the optimization results, the core outlet temperature distribution is changing with the elevation angle of the reactor system. The pancake core type has larger temperature distribution change as the elevation angle changes due to the sea wave. The natural circulation capability is good for safety. However, large driving head of natural circulation may cause large temperature fluctuation as the elevation angle changes.     

Consequence calculation for relocation after nuclear accidents

This paper presents the results of dose and cost calculation for relocation after nuclear accidents. In order to quantify the relationship between radiation dose and relevant parameters defining protective actions, well-designed calculations have been performed and the results analyzed. On the basis of this, some dependencies between the important parameters describing the features of relocation and decontamination have been mathematically formulated. The similar has also been done for some economic costs as a result of implementing the countermeasures considered in this paper.     

Transient behaviour of a parallel pump in nuclear research reactors

A theoretical analysis of the fast transients of a parallel pump, based on inertia of the rotating parts and inertia of the fluid, is proposed. It leads to total torque, total head, and total system resistance during transient periods. The equations indicate that an increase in coolant inertia increases the acceleration head. While an increase in the moment of inertia of rotating parts decreases the acceleration head. The model is used to analyze the behaviour of the Tehran Research Reactor (TRR) primary coolant loop parallel pump during a fast startup. The results of present model are compared with similar studies and good agreement is observed.     

Supercritical-pressure, light-water-cooled reactors for economical nuclear power plants

Design studies of supercritical-pressure light-water-cooled reactors (SCLWRs) have been carried out to pursue drastic improvement of the economy of nuclear power generation. The core is cooled by supercritical water which is superheated without the phase change. The cooling system is a once-through type; the whole core flow is driven by the feedwater pumps and is directly led to the turbine. No recirculation line is necessary. Besides, steam separators and dryers are not needed. Water rods are used to enhance the moderation and to increase the flow velocity around the fuel rods. The radial peaking factor is satisfactorily reduced by controlling uranium enrichment and gadolinia concentration as well as water rods. Flattening of the radial power distribution is important to enhance the thermal efficiency. This can be achieved by the coolant density feedback and the out-in refueling pattern. Orificing is also effective to enhance the thermal efficiency. The thermal efficiency is above 40% with stainless steel cladding. Plant control system and safety system are also designed. The core flow should be directly maintained due to the once-through direct cycle. Plant behaviors of large break LOCAs and loss of offsite power are analyzed. Safety criteria are satisfied in both cases. The feasibility of SCLWR is shown.     

Improvements and validation of the transient analysis code MOREL for molten salt reactors

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.     

秦山核电厂在假想失水事故工况下放射性碘的辐照后果预测

本文根据压水堆假想失水事故工况下放射性碘在安全壳中的转移和向大气环境扩散的计算模型,编制了计算程序SRIC,并用该程序对秦山核电厂最大假想失水事故短期内由放射性碘所引起的辐照后果进行了预测计算。     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

秦山核电厂全厂断电事故厂外后果分析

采用MELCOR和MACCS程序对秦山核电厂全厂断电事故的源项和厂外后果进行了计算。该事故会引起厂外群体受到较大剂量的放射性照射，但剂量不足以引发早期确定性健康效应。并对可能采取的应急防护行动进行评估，确定最佳防护措施为：安全壳泄漏阶段实施隐蔽；若安全壳超压失效无法避免，应急计划区内应立即实施撤离。     

Requirements for D-D tandem mirror reactors

Requirements for D-D barrier tandem mirror reactors are calculated from an equilibrium power balance model. To obtain adequate plasmaQ and reasonable power density, axisymmetric configurations are required to decrease barrier length and radial transport and to increase central cell beta. We find that for a reactor producing 900 MW net electric power from aQ=6.5 plasma, a central cell length of 225 m, maximumB of 15 T, and neutral beam injection energy of 700 keV are necessary. In addition to high central cell beta (70%), high barrier beta (40%) is needed to allow the ECRH power required to reduce the barrier potential. Using too much barrier ECRH power results in a decrease inQ. Nuclear elastic scattering of fusion products plays an important role in the overall plasma power balance. When nuclear scattering and coulomb scattering are included, the plasmaQ value is increased by more than 40% compared to the case when coulomb scattering alone is considered.     

Robust nonlinear control for nuclear reactors using sliding mode observer to estimate the xenon concentration

This paper presents findings on the sliding mode controller for a nuclear reactor. One of the important operations in nuclear power plants is load following. In this paper, a sliding mode control system, which is a robust nonlinear controller, is designed to control the pressurizedwater reactor power. The reactor core is simulated based on the point kinetics equations and six delayed neutron groups. Considering neutron absorber poisons and regarding the limitations of the xenon concentration measurement, a sliding mode observer is designed to estimate its value, and finally, a sliding mode control based on the sliding mode observer is presented to control the core power of reactor. The stability analysis is given by means Lyapunov approach; thus, the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications, and moreover,the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed observerbased controller in terms of performance, robustness and stability.     

Applicability of 100 TBq cesium 137 release into environment as a safety criterion for consequence assessment at reactor design approval stage

The International Atomic Energy Agency (IAEA) fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. Therefore, a severe accident consequence assessment has to be able to include all quantifiable consequences on people and the environment. Our previous studies on estimation of cost per severe accident succeeded in quantifying aforementioned consequences. However, the estimation requires enormous quantity of data, time and human resources, thus it may be inappropriate at the reactor design approval stage. Finnish government uses “100 TBq cesium 137 release into environment”, which was proved to generate limited health effects, as one of the reactor design criteria for accident consequences. In this study, we perform an evaluation of annual dose from the 100 TBq cesium 137 release and confirm limited health effects. We form the environmental impact index based on insights from our previous studies and used it to assess consequences to the environment. The estimated environmental impact index is very small, which confirms the limitedness of the environmental impacts of the release. These findings ensure the applicability of 100 TBq cesium 137 release into environment as a safety criterion for consequence assessment at reactor design approval stage.     

Innovative nuclear technology based on modular multi-purpose lead–bismuth cooled fast reactors

Today's nuclear power is in the state of an intrinsic conflict between economic and safety requirements. This fact makes difficult its sustainable development.

One of the ways of finding the solution to the problem can be the use of modular fast reactors SVBR-75/100 cooled by lead–bismuth coolant that has been mastered in conditions of operating reactors of Russian nuclear submarines.

The inherent self-protection and passive safety properties are peculiar to that reactor due to physical features of small power fast reactors (100 MWe), chemical inertness and high boiling point of lead–bismuth coolant, integral design of the pool type primary circuit equipment.

Due to small power of the reactor, it is possible to fabricate the whole reactor at the factory and to deliver it to the NPP site in practical readiness by using any kind of transport including the railway.

Substantiation of the high level of reactor safety, adaptability of the SVBR-75/100 reactor relative to the fuel type and fuel cycle, issues of non-proliferation of nuclear fissile materials, opportunities of multi-purpose usage of the standard SVBR-75/100 reactors have been viewed in the paper.     

An improved semi-implicit direct kinetics method for transient analysis of nuclear reactors

Semi-implicit direct kinetics(SIDK) is an innovative method for the temporal discretization of neutronic equations proposed by J. Banfield. The key approximation of the SIDK method is to substitute a timeaveraged quantity for the fission source term in the delayed neutron differential equations. Hence, these equations are decoupled from prompt neutron equations and an explicit analytical representation of precursor groups is obtained,which leads to a significant reduction in computational cost. As the fission source is not known in a time step, the original study suggested using a constant quantity pertaining to the previous time step for this purpose, and a reduction in the size of the time step was proposed to lessen the imposed errors. However, this remedy notably diminishes the main advantage of the SIDK method. We discerned that if the original method is properly introduced into the algorithm of the point-implicit solver along with some modifications, the mentioned drawbacks will be mitigated adequately. To test this idea, a novel multigroup, multi-dimensional diffusion code using the finitevolume method and a point-implicit solver is developed which works in both transient and steady states. In addition to the SIDK, two other kinetic methods, i.e., direct kinetics and higher-order backward discretization, are programmed into the diffusion code for comparison with the proposed model. The final code is tested at different conditions of two well-known transient benchmark problems. Results indicate that while the accuracy of the improved SIDK is closely comparable with the best available kinetic methods, it reduces the total time required for computation by up to 24%.     

核电厂风险管理活动中的PRA质量要求

基于美国核管理委员会(NRC)推行的在核电厂运用的概率安全评价(PRA)技术,介绍PRA质量的含义、NRC在应用PRA过程中提出的分阶段提高PRA质量的方法以及相应的管理导则。结合国内现状,给出提高PRA质量的可接受方法。