Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity (ρ), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.     

原型微堆低浓化初步研究

利用蒙特卡罗计算程序,对高浓铀为燃料的原型微堆的有效增殖因数、控制棒价值、上铍反射层价值以及辐照座内的中子注量率等参数进行了计算。将计算值与实验结果进行了比较,两者基本相符。在原型微堆堆芯尺寸保持不变的情况下,将堆芯燃料元件芯体用富集度为12.5%UO₂替换UAl和用锆包壳替换铝包壳,对堆芯燃料低浓化方案进行了计算,给出了方案的计算结果。并利用RELAP5程序计算了原型微堆低浓铀堆芯阶跃引入4.0 mk反应性情况下反应堆的相关参数。     

混合能谱超临界水冷堆反应性引入事故安全分析

本文研究了混合能谱超临界水冷堆(SCWR-M)在发生控制棒失控提升事故和弹棒事故这两类反应性引入事故后的反应堆系统响应。首先利用修改的可用于超临界条件下的系统程序RELAP5对混合能谱超临界水冷堆进行系统建模,并计算分析在功率运行工况下事故过程中功率、流量及包壳温度等重要参数的变化趋势,最后对反应性参数如控制棒价值、控制棒抽出速率和负反馈系数进行了参数效应分析。结果表明,在设计工况下混合能谱超临界水冷堆系统可有效地将衰变热导出堆芯,保证了燃料棒的完整性。另外,反应性参数对控制棒失控提升事故的安全性影响不大,但对弹棒事故的包壳峰值温度影响很大,过于保守的反应性参数估计会使安全裕量大为减小。     

Parametric tests and measurements after shimming of a beryllium reflector in a miniature neutron source reactor (MNSR)

After 10 years operation of Pakistan research reactor-2 (PARR-2), a miniature neutron source reactor (MNSR), a beryllium reflector was added to compensate the loss of reactivity due to burn up of fuel. Beryllium shim plates have been placed at the top of the core in a tray provided for this purpose. The control rod was dismantled and withdrawn from the core and the reactor was made subcritical with cadmium shimming. To monitor the neutron population during this experiment, two additional neutron monitoring channels based on BF₃ were installed around the core. Measurement of important Parameters such as effective delayed neutron fraction, decay constant, excess reactivity, control rod worth, temperature coefficient of reactivity, thermal neutron flux, cadmium ratio was done after the addition of Be reflector. Increase in reactivity worth due to addition of Be shim was 1.0 mk.     

A model for uniform Zircaloy clad corrosion in pressurized water reactors

A phenomenological water-side corrosion model for Zircaloy fuel cladding for pressurized water reactors (PWRs) is considered. The model acounts for the breakaway transition in the Zircaloy oxidation rate that takes place in an isothermal condition and the changes that occur during reactor operation, i.e. the dependence of oxide growth on fast neutron flux and cladding oxide layer thickness. Closed-form analytical solutions of the oxidation kinetics equations are obtained. The corrosion kinetics model is coupled to PWR thermal and hydraulic models which assume a subchannel that is either a closed single channel or a multichannel which accounts for coolant cross-flow and coolant enthalpy mixing. Both single-phase forced convection and subcooled nucleate boiling are accounted for in the thermal-hydraulic models. The model calculates the coolant temperature at the axial midplane of each axial segment of the fuel rod. When an oxide layer is present, the temperature at the metal-oxide interface is determined. This temperature in turn is used to determine the oxide growth via the Arrhenius temperature dependence of the Zircaloy oxidation rate. The predictions of the model have been compared with the measured cladding oxide data obtained in PWRs. The data for a given rod were obtained at various burn-ups (at the end of reactor cycles) and various axial positions of the rod. Our evaluations show that the model predicts the measured data satisfactorily; however, the deviations are discussed. The model has been used to study the effect of core loading patterns on cladding oxide growth. Our analyses show that core nuclear design is an important factor for water-side corrosion of fuel rods.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

CSR1000控制棒落入堆芯三维瞬态分析

针对超临界水堆(SCWR)控制棒落入堆芯事件特点,采用堆芯三维瞬态性能分析方法,利用开发的SCWR堆芯三维瞬态物理-热工水力耦合程序STTA,建立SCWR堆芯落棒瞬态三维计算模型和分析流程,研究分析超临界水堆CSR1000在控制棒落入堆芯瞬态过程中的堆芯性能,分析评价落棒瞬态下CSR1000堆芯的安全性能。堆芯三维落棒瞬态分析表明,当落入堆芯棒束价值较高时,落棒初期堆芯功率下降较快,之后由于水密度的反应性反馈,堆芯功率缓慢回升至新的平衡,堆芯功率下降速率超过了停堆信号整定值,将触发保护停堆;当落入堆芯棒束价值较低时,由于水密度的反应性反馈,堆芯功率下降缓慢,堆芯功率下降速率未能达到停堆信号整定值,不能触发保护停堆。控制棒落入堆芯对堆芯轴向功率分布影响很小,高价值落棒导致的落棒区域燃料组件功率坍塌相对低价值落棒更明显。无论是高价值落棒还是低价值落棒,瞬态过程中最大包壳壁面温度均低于瞬态安全限值850℃。水密度的显著反应性反馈及必要的保护停堆措施能保证CSR1000堆芯在控制棒落入堆芯过程中的安全性能。     

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρ_ex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU).     

Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling

Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650 °C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6 mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156 W/cm³ and its equivalent diameter is only 2.7 m for 1000 MWe class reactor core. High average core outlet temperature of 500 °C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost.     

Heat transfer to liquid metals flowing turbulently and longitudinally through closely spaced rod bundles : Part I

This paper presents results of a theoretical study of heat transfer to liquid metals in fully developed turbulent, in-line flow through unbaffled, spacer-free rod bundles. The bundles have equilateral triangular arrangement; and the rod spacings, rod design, and ranges of independent variables covered were chosen with reference to liquid-metal-cooled nuclear reactor applications. Three different sets of thermal boundary conditions are considered: (A) uniform heat flux in the axial direction with uniform temperature in the circumferential direction, on the outer surface of the cladding; (B) uniform heat flux in both directions, on the outer surface of the cladding; and (C) uniform heat flux in both directions on the inner surface of the cladding. The results of the third set are presented in Part II.     

Radiation Effects Problems in Nuclear Fuel Rods

This paper will review some of the radiation effects problems that have been encountered in nuclear fuel rods. Examples are drawn from both the commercial Light Water Reactors (LWR) and the new, developing Liquid Metal Fast Breeder Reactors (LMFBR). The performance of a particular fuel rod is influenced by (1) the interaction between the ceramic fuel pellet column and the metal cladding tube, and (2) the interactions of the fuel cladding, separately, with the reactor core environment. The latter includes the effect of neutron flux, temperature, stress and corrosive chemicals. Despite this complex and hostile environment, many of the material interactions are known. Sufficient new detail will enable the fuel rod and the core power operation to be designed so that a very low incidence of fuel rod failures can be maintained.     

医院中子照射器反应堆实验研究

医院中子照射器是专用于硼中子俘获治疗的核装置,所用反应堆功率为30 kW,采用~(235)U富集度为12.5%的UO_2为燃料,金属铍反射层,轻水为慢化剂和冷却剂.堆芯产生的热量靠自然循环冷却.在反应堆堆芯相对两侧分别设置了热中子束流和超热中子束流,用于治疗患者.在微堆零功率实验装置上,完成了临界质量、控制棒效率、上铍反射层效率及其它部件反应性的测量,确定了最终燃料元件的装载,为工程物理启动提供实验数据.     

超临界水冷堆燃料组件及堆芯方案简化设计研究

为解决超临界水冷堆中子慢化不足的问题,采用在燃料组件中设置“水棒”或者加入固体慢化剂的设计方法,同时堆芯冷却剂采用多流程流动方案,导致燃料组件和堆芯结构复杂化,并向堆内引入较多强中子吸收结构材料。因而基于CSR1000研究结果,开展了简化超临界水冷堆燃料组件及堆芯结构设计。研究结果有效简化了超临界水冷堆燃料组件和堆芯结构。      

Neutron distribution influence on the thermal design of a nuclear fuel rod

The object of the present paper is to look into the thermal and thermoelastic state in a nuclear fuel rod with gap and cladding, including a neutron flux distribution deduced by diffusion theory. The steady state temperature and the thermal stresses in the rod are analytically determined solving a set of differential and integral equations, deriving from heat conduction and Hooke's law. The mathematical technique here utilized is based on the modified first kind Bessel functions theory; the solution presents a very compact and simple form.

The results so obtained are related and a comparison is made with the analogous results deduced implying a constant heat source distribution in the fuel, i.e., neglecting the actual neutron flux distribution. Some numerical results are finally reported with reference to a common boiling water reactor and shortly discussed.     

Critical neutron heating in the control drums of a dual purpose thermionic space reactor for power and propulsion

The critical neutron heating in the reflector control drums is investigated for a fast incore thermionic space craft reactor for power and nuclear propulsion. The reactor is fueled with uranium carbide (UC) and controlled with the help of rotating B₄C drums imbedded into the beryllium reflector. While the neutron heating in the drums would not require a cooling mechanism in the power phase, the heat generation during the thrust phase obliges cooling for a nuclear thermal thrust around F = 5000 N by a specific impulse of 670 s⁻¹ at an hydrogen exit temperature around 1900°K. With a beryllium reflector without extra cooling measures, thermal thrust must be kept F < 2500 N to relieve the thermal load in the reflector. On the other hand, a reflector made of BeO may withstand a thermal load for a nuclear thermal thrust of F = 5000 N. The neutronic analysis has been conducted in S₁₆-P₃ and S₈-P₃ approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. A reactor control with boronated reflector drums (drum diameter = 14 cm) at the outer periphery of the radial reflector of 16 cm thickness would make possible reactivity changes of Δk_{eff = 13.55%}—amply sufficient for a fast reactor—without a significant distortion of the fission power profile during all phases of the space mission. Calculations are conducted for a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels around 50 kW_el.     

FARST — A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:

1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.

2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.

3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.

4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.

The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR).The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant.     

Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k_eff core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape.     

Development of Hydride Absorber for Fast Reactor—Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B₄C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH_1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.     

Evaluation of Gap Heat Transfer between Boron Carbide Pellet and Cladding in Control Rod of FBR

Heat transfer across a gas gap between a boron carbide pellet and a cladding in FBR control rod has been experimentally investigated. The main purpose of this investigation is to present a calculational method for the gap heat transfer in order to improve accuracy of thermal design for the control rod with 0.5 mm gap width at a beginning of reactor operation.

Two types of tests have been carried out using simulated control rods. One is low temperature tests below 200°C. The test results indicated that the convective heat transfer has a negligible effect on the gap conductance when the Rayleigh number using the gap width as the characteristic length is below 0.1. The other is high temperature tests up to 600°C.

The results showed +10 to — 5% variations in the gap conductance data due to eccentricity between the pellet and the cladding. The prediction based on conduction and radiation heat transfers considering a thermal expansion and an eccentricity gave better results of gap conductance having a maximum difference of only 17% from the measured ones. Calculation in the radiation heat transfer used thermal emissivities. 0.85 for the boron carbide and 0.15 for the cladding, measured by infrared thermography.     

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA FRTMB分析了CAP1400反应堆直接注射（DVI）管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。