燃耗信任制分析方法在乏燃料装卸与贮存系统中的应用

介绍了在乏燃料系统中应用燃耗信任制分析方法的一般过程.提出了对验证所使用计算程序的要求.阐述了燃耗信任制分析过程中的包络性原则与保守性原则.     

乏燃料贮存燃耗信任制计算的敏感元素分析

燃耗信任制的应用分析需要计算乏燃料系统的反应性,为了保证临界安全需要保证分析结果的包络性,为此需要考虑到各种不同燃耗因素对于分析结果的影响。采用SCALE程序包中的STARBUCS模块,通过对OECD/NEA发布的Phase-IA及Phase-IIA基准题的验证分析,研究了不同信任等级、堆芯运行参数等因子对乏燃料贮存系统临界安全的敏感性分析,并得出有益的结论。     

燃耗信任制的应用方法

在临界安全分析中采用燃耗信任制可在确保安全性的基础上带来巨大的经济效益。本文详细介绍了当前燃耗信任制的应用方法,如何确定所考虑的核素种类和确定燃耗计算状态得出保守的核素核子密度以及如何确定轴向燃耗分布不均匀的组件的置信燃耗深度,最后将燃耗信任制应用于某乏燃料溶解器,比较了此方法带来的效果。     

燃耗信任制临界计算中保守性因素研究

在运用燃耗信任制技术进行乏燃料储存、运输等环节的临界安全分析时,临界计算所采用的条件是否具有足够的包络性十分关键。本文借助于OECD/NEA发布的若干燃耗信任制临界安全基准题,使用SCALE5.1软件中的STARBUCS模块进行分析,对信任核素选取、乏燃料冷却时间以及端末效应等因素对乏燃料系统临界安全性的影响进行了研究,得出了各参数保守性的有关结论。     

燃耗信任在PWR乏燃料贮存水池临界安全设计中的应用和发展综述

本文述评了燃耗信任在PWR乏燃料贮存水池核临界安全设计中的应用和发展现状，主要包括燃耗信任对临界应用的原理、规范和标准现状、国外发展动态，燃耗信任水平，信任同位素组选择，辐照历史、运行条件和冷却时间的考虑，轴向燃耗分布的末端效应，装载曲线和燃耗验证等．     

基于计算关系曲线的无源中子燃耗测量方法的研究

采用燃耗信任制技术可显著提高乏燃料贮存及运输的经济性,也是国际上该领域的发展趋势。非破坏性燃耗测量是采用燃耗信任制技术必须解决的关键问题之一。在非破坏性燃耗测量方法中,基于计算关系曲线的无源中子燃耗测量方法可以精确地测量乏燃料组件的平均燃耗,结合总γ方法,还可以测量出乏燃料组件的末端燃耗。根据该方法的基本原理,在调研分析的基础上,确定了燃耗测量分析方法及其流程。其中,确定乏燃料燃耗与中子发射强度之间的关系、中子发射强度计算方法以及Keff的快速计算方法是测量分析方法的关键技术。     

田湾核电站乏燃料水池采用燃耗信任制的计算研究

以田湾核电站(TNPS)2×5排列的贮存格架构成的乏燃料水池为例,研究采用燃耗信任制技术的密集贮存和临界安全问题。采用MONK9A程序计算分析不同富集度、不同燃耗的乏燃料装载情况下系统的k_eff. 根据系统k_eff随不同初始富集度燃料的燃耗变化情况给出了水池的参考装载曲线。采用燃耗信任制技术的密集贮存方案能提高贮存能力31%。     

AP1000乏燃料贮存格架临界安全分析

基于SCALE6程序包对西屋公司采用燃耗信任制技术的AP1000核电厂乏燃料贮存格架(SFSRs)临界安全分析过程进行了复现,在此基础上结合AP1000核电厂堆芯反应性控制特性,分析了轴向燃耗分布对系统反应性的影响。结果表明,高燃耗下采用机械补偿(MSHIM)轴向燃耗分布计算得到的系统反应性更保守,同时临界安全分析中需考虑吸收体在组件燃耗过程中对反应性的影响,且不应信任可溶硼。     

基于燃耗信任制的核电厂乏燃料贮存水池临界计算

为研究初始富集度为4.95%的新型燃料组件卸料后高密度贮存的可行性,以岭澳核电站3、4号机组乏燃料贮存水池为例,利用SCALE5.1程序系统中基于燃耗信任制的STARBUCS临界计算程序,分析了该新型燃料组件在不同燃耗情况下,锕系核素和裂变产物的产额变化及其对反应性的影响;基于锕系加裂变产物信任水平,计算了燃料组件在不同燃耗深度和不同贮存年限情况下的乏燃料贮存水池临界安全性;给出了乏燃料贮存水池Ⅱ区的参考装载曲线。计算表明:该新型燃料组件在燃耗达到45 GWd.t-1(U)后可以高密度贮存在乏燃料贮存水池Ⅱ区。     

田湾核电站乏燃料干式贮存临界安全计算研究

以CASTOR 1000/19干式贮存容器装载田湾核电站六角形乏燃料组件为例,研究六角形乏燃料干式贮存的临界安全问题。基于新燃料假设,应用MONK9A程序对贮存容器满装载乏燃料进行不同工况下k_eff的计算。计算结果表明：正常工况下,k_eff远小于临界安全限值,是临界安全的;事故工况下,当²³⁵U富集度大于3.15%时,系统存在临界安全风险,须减少乏燃料装载量来确保临界安全。考虑燃耗信任制后,采用相同的模型计算得出贮存容器满装载的参考装载曲线,按此曲线要求装载能确保所有工况下的系统临界安全。采用燃耗信任制技术提高了贮存容器的利用率。该研究可为田湾核电站采用乏燃料干式贮存方案提供依据。     

燃耗数据库基准检验方法研究

燃耗数据库基准检验方法对于研制高准确度的燃耗数据库至关重要。本文以TAKAHAMA 3压水堆辐照后检验实验中SF95样品的建模为例，研究了建模要素对燃耗计算的影响，确定了燃耗实验建模的方法，开展了燃耗信用制研究感兴趣的锕系和裂变产物核素积存量计算值与实验值的比对。比对结果显示，主锕系核素计算偏差小于2%，大部分次锕系核素偏差小于10%，大部分重要裂变产物核素偏差小于5%。本文还对125Sb积存量随燃耗深度变化规律进行了理论分析，确认了破坏性放化实验测量结果存在缺陷，并进一步获得了125Sb积存量的修正值，使计算偏差从接近170%下降到20%以内。本次研究表明，燃耗数据库基准检验研究不仅需发展适当的燃耗实验建模方法，还需对实验数据进行适当的评价。     

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of ²³⁷Np, ²³⁸Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of ²³⁷Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO₂ or UO₂–Gd₂O₃), which was similar to the case of PWR fuel. The differences in the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ¹⁴⁹Sm, and ¹⁵¹Sm have a large impact on k_eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.     

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

Fuel burnup performance of an OTTO refueling pebble bed reactor with burnable poison loading

Fuel burnup performance has been analyzed for a pebble bed reactor with a once-through-then-out (OTTO) refueling scheme and compared with a reference multi-pass scheme. A new fuel pebble was designed by adding spherical B₄C particles into its free fuel zone for controlling the infinite multiplication factor during burnup, and then reducing the axial power peak of the OTTO scheme. The objective is to maximize the fuel burnup performance of the OTTO scheme while keeping the power peak under a limit and ensuring the core criticality. Numerical calculations were performed based on the 400 MWt pebble bed modular reactor (PBMR) using the MVP code. For the fuel pebble of the PBMR containing 9 g uranium with 9.6 wt% ²³⁵U enrichment, 1600 B₄C particles with a radius of 70 μm are determined to flatten the k_∞ curve in the early burnup stage. The dependences of the neutronic properties of the core with the OTTO scheme on target fuel burnup show that the maximum target burnup of 74 GWd/t can be achieved so that the power peak is reduced to about 10.80 W/cm³ which is approximate that of the multi-pass scheme (10.85 W/cm³). This target burnup is about 22% less than that of the multi-pass scheme (95 GWd/t), i.e. the fuel utilization efficiency of the OTTO scheme is about 22% lower, which could be compensated by the construction and operation cost of the fuel handling system. This result also suggests that further investigations of the fuel burnup performance and other properties are needed in both neutronic and thermal hydraulic viewpoints to find out the optimal core performance.     

Impact of cross section libraries on CASMO-4E and CASMO-4 results

A detailed comparison between the code CASMO-4 and its extended version CASMO-4E has been made. In addition to the standard library, CASMO-4E calculations have been performed also with its extended libraries. The differences are significant enough to be considered when choosing the library to be used for a particular problem. The differences in the multiplication factor k_∞ range up to several hundred pcm depending on the void history, burnup and other parameters. The differences in fuel temperature or void coefficients are smaller especially at small void fraction and low burnup. At large void and low burnup CASMO-4E with the standard library gives significantly different results than the other combinations. The microscopic cross sections show small differences when calculated with the same library but clear differences due to the extended libraries.     

An optimal loading principle of burnable poisons for an OTTO refueling scheme in pebble bed HTGR cores

An optimal loading principle of burnable poisons (BPs) is proposed to eliminate the problem of an excessively high power peaking factor in once-through-then-out (OTTO) pebble bed HTGR cores. The effectiveness of the principle is confirmed through neutronic and thermal hydraulic calculations for an HTGR core with thermal power of 600 MWt. Spherical BP particles are distributed uniformly together with TRISO-coated fuel particle inside the free fuel zones of fuel pebbles to maintain constant k_∞ during burnup. The applicability of BP materials, such as B₄C, Gd₂O₃, Sm₂O₃, Eu₂O₃, Er₂O₃, CdO, and HfO₂, is investigated. Because complete BP depletion at the target burnup is required to avoid reactivity loss and extra enrichment, the four BP materials considered as applicable candidates are B₄C, Gd₂O₃, Er₂O₃, and CdO. The complete BP depletion is demonstrated by a burning fraction at the target burnup. The power density is excessively high at the core top and the power peaking factor is 4.44 when no BP is added to the fuel pebbles. Optimal BP loading reduces the power peaking factor from 4.44 to about 1.7. Because of the power peaking factor reduction, the maximum fuel temperatures are lower than the maximum permissible values of 1250 °C for normal operation and 1600 °C during a depressurization accident.     

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

In the design of fast reactor core with higher burnup and higher linear power, prediction accuracy of burnup history of fuel pin should be upgraded so as to assure fuel integrity without extra design margin under increased neutron fluence and burnup. A method is studied to predict fuel pin-wise power and its burnup history in fast reactors accurately based on an analytic solution of diffusion theory equation on hexagonal geometry with boundary condition from core calculation by finite-differenced diffusion calculation code. The present method is applied to a fast reactor core model, and its accuracy in predicting fuel pin power is tested. The result is compared with the reference solution by the finite difference calculation with very fine mesh. It is found that the present method predicts the power peaking factors in fuel assemblies accurately. The fuel pin-wise nuclide depletion calculation is also done using neutron fluxes for each fuel pin. The result shows that the fuel pin-wise depletion calculation is very important in predicting the burnup history of the fuel assembly in detail.     

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of ²³⁵U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k _eff's of the critical cores were about ?1:2%Δk for the diffusion calculations (JENDL-3.2), ?0:5%Δk for the transport calculations (JENDL-3.3), and ?0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was ?2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.     

Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron cross-section libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged.     

Problems and Experience in Guaranteeing the Nuclear Safety of the Storage of Spent Fuel from Nuclear Power Plants

Nuclear-safety problems are examined and the results of investigations of nuclear safety of storage sites are presented for spent nuclear fuel from nuclear power plants. The initial events of anticipated and unanticipated accidents, methods and errors in the calculation of k _eff taking account of burnup to ensure nuclear safety, the possibility of measuring k _eff of storage sites experimentally, and new forms of fuel with a consummable absorber are calculated.