Fission Product Model for BWR Analysis with Improved Accuracy in High Burnup

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation.     

Measurement of Burnup in FBR MOX Fuel Irradiated to High Burnup

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system.

Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector.     

Prediction of Nongaseous Fission Products Behavior in the Primary Cooling System of High Temperature Gas-Cooled Reactor

In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified.

This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model.     

Fission Products Release from Irradiated FBR MOX Fuel during Transient Conditions

Out-of-pile experiments on the release of fission products (FPs) under transient heating conditions were carried out for mixed oxide fuels. The fuels used in the experiments, the plutonium content of which was 30 wt%, were irradiated up to 65 GWd/t in the experimental fast reactor JOYO. The experiments consisted of two runs, FP-1 and FP-2. In FP-1, the fuel sample was first heated to 2,000°C and then up to 3,000°C. The holding time was 30 min at each temperature. In FP-2, the terminal temperatures were 1,500°C and 2,500°C, and the holding time was 30 min in the same manner.

The release ofCs, a volatile FP, was detected as soon as the fuel sample was heated up. The release rate, after peaking in several minutes, decreased gradually via a diffusion process in the fuel matrix. The “gross” diffusion coefficient agreed well with the data reported in other experiments using LWR fuels. The release fractions were identical in both experiments, namely Cs ～100%, Sb ～100%, Ru ～10%, Ce ～0% and Eu ～0%.     

Dependence of Fast Reactor Fuel Burnup Characteristics on Nuclear Data Libraries

In the framework of the development of burnup calculation method for commercial fast reactors, a sensitivity analysis was carried out to clarify the dependence of fuel burnup characteristics on nuclear data libraries (NDLs). The following NDLs were compared: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2 and JENDL-3.3. The NDL-dependence of material balance for main heavy metal nuclides (²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu) was small, since the number densities at the end of one-cycle burnup did not change over 1 or 2% among the above-mentioned libraries. Relatively large differences were found for minor actinide nuclides, especially for ²³⁶U, ²³⁷Np, ^242mAm, ²⁴³Am and curium isotopes. The number densities for these nuclides after burning up showed remarkable NDL-dependence over 5% through 50%. A burnup sensitivity analysis system based on the generalized perturbation theory enabled us to find out quantitatively the causative nuclides and reactions, as well as their energy regions.     

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV) of 21:4 _ 1026 and 23:8 _ 1026 n/m2, without any indication of fuel pin breaching. Structural components of these assemblies were made of modified type 316 stainless steel and 15Cr-20Ni base advanced austenitic stainless steel. Postirradiation examinations of these assemblies revealed dimensional changes of fuel pins and assembly ducts due to irradiation-induced void swelling and irradiation creep, and fuel cladding local oval distortions due to bundle-duct interaction (BDI). The swelling resistance of 15Cr-20Ni base advanced austenitic stainless steel fuel pin cladding was almost the same as that of the modified type 316 stainless steel cladding, while the assembly duct of the former material had a slightly higher swelling resistance than that of the latter material. Analyses of fuel pin bundle deformations indicated that these assemblies likely mitigate BDI mainly by fuel pin bowings and cladding oval distortions.     

Fission Density in Plutonium-Uranium Mixed Oxide Fuel Irradiated by Pulsed Power in NSRR Experiments

Fuel behavior during a reactivity initiated accident condition is recognized to be predo minantly related to energy deposition in the fuel. The first stage of NSRR in-pile experiments addressing the behavior of PuO₂-UO₂ mixed oxide fuels evaluated the energy deposition per unit integrated reactor power by γ-ray spectrometry. Solid samples were used to measure the γ-rays because the facility is permitted to handle solid plutonium only. Determination of the penetration ratios of γ-rays from the actinides contained in the fuels allowed correction for the self-attenuation of γ-rays in the solid samples. Evaluation of the effect of epithermal neutron fissions was also necessary since the fissile nuclides of ²⁴¹Pu and ²³⁹Pu have high resonance cross sections in the epithermal energy region. For this evaluation, the fission density was first calculated for the fission products as a function of the contribution ratio of the fissions of epithermal neutrons. Accurate fission density was then determined using the contribution ratio which minimized the deviation of the calculated values for the fission density. The fission densities determined by this simplified method agreed well with the values calculated using the computer codes CITATION and GGC-4.     

Melting Temperature of Irradiated UO2 and UO2-2wt%Gd2O3 Fuel Pellets up to Burnup of about 30 GWd/tU

Melting temperature of UO₂ and UO₂-2w/oGd₂O₃ fuels irradiated in a commercial LWR were determined by a thermal arrest technique in a burnup range up to approximately 30GWd/tU.

No decrease in the melting temperature was observed on both UO₂ and UO₂-2w/oGd₂O₃ fuels with increment of burnup to 30GWd/tU. It was also found that the Gd₂O₃ addition below 2w/o has no influence on the melting temperature.     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

Oxygen Potentials of UO2 Fuel Simulating High Burnup

The oxygen potentials at 1,000 and 1,300°C and the lattice parameters of UO₂ fuels with soluble fission product elements (Zr, Ce, Pr, Nd, Y), simulating high burnup of up to 10^a,_o have been measured by means of thermogravimetry and X-ray diffraction. The oxygen potentials for (U, FP)O₂+x fuels are higher than pure UO_2+x; at a given composition and increase positively with increasing simulated burnup. They can be represented as a function of the mean uranium valence at compositions of 0/M>2.01. The lattice parameters of stoichiometric (U, FP)0₂.00 fuels decrease linearly with simulated burnup, and can be expressed as a (pm) = 547.02–0.1225, where B is burnup in ^a._o     

Feasibility of Fast Fission System Confining Long-Lived Nuclides

The feasibility of fast fission system confining long-lived nuclides without other supporting system as synergetics for fuel sustainment and waste incineration was studied from the aspects of nuclear material balance and neutron economy. The continuous utilization of fast fission system which confines all actinides in the reactor but discharges all FP will lead to huge accumulation of radioactive wastes such as ¹²⁹I, ¹³⁵Cs, ¹⁰⁷Pd, ⁹³Zr, ⁹⁹Tc, ¹²⁶Sn and ⁷⁹Se in the far future. Then we studied the feasibility of the system that these long-lived seven FP are also confined in the reactor with actinides. In this scheme, all the long-lived nuclides to be disposed of were exposed with neutrons in the reactor and removed as different nuclides after nuclear transmutation. As the wastes stored in the repository was composed of only shorter-lived nuclides, total amount of radioactive wastes in the repository was suppressed to be less than a few tons per 3 GWt reactor.     

Sequential Ion-Exchange Separation of Heavy Elements and Selected Fission Products for Burnup Measurement

A sequential ion-exchange separation method was developed for use in burnup measurements of nuclear fuels. Group separation by anion-exchange resin column with hydrochloric acid solutions containing small amounts of nitric acid and hydrochloric acid was followed by various cation and anion- exchange processes. The heavy elements, such as U, Np and Pu, and some fission products selected as burnup monitors, such as Cs, Mo and Nd, could be sequentially and quantitatively separated from a sample taken from spent fuel. The recovery of these elements through the separation processes were examined. The sampling ratio of an aliquot in reference to the whole fuel specimen was determined by adding as sampling monitor a known amount of Cu to the sample during dissolution. The validity of the ion-exchange separation technique for routine analysis for burnup measurements is also discussed.     

高燃耗陶瓷燃料颗粒的参数化建模及强度分析

弥散燃料芯体中的陶瓷燃料颗粒在辐照条件下会形成裂变气孔,燃料颗粒内部气孔间的相互干涉作用及气孔内压的增长致使局部拉应力超过材料强度极限,进而导致燃料颗粒开裂。本文考虑高燃耗燃料颗粒内气孔尺寸和位置分布的非均匀性,实现了颗粒内部的细观结构参数化建模。运用有限元方法计算并分析了气孔尺寸、基体约束压应力、温度和气孔分布方式对颗粒内部最大拉应力的影响,研究了颗粒内开裂危险区的分布规律。结果表明,陶瓷燃料颗粒最大拉应力随气孔尺寸和温度的增加而增大,随基体约束压应力的增加而减小;燃料相的断裂强度减小,开裂危险区面积增大;燃料颗粒从内部多处开裂破坏,而表层处开裂的概率更大。本文为弥散燃料失效研究及优化设计提供了分析方法及数值参考。     

Post Irradiation Examination of High Burnup PWR Fuel

Post irradiation examination (PIE) of a high burnt lead fuel assembly, which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GWd/t maximum fuel assembly burnup.

The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NFI) and achieved the maximum burnup of 45 GWd/t after 4 cycles of irradiation in Ohi Unit 1 of the Kansai Electric Power Co. Inc. (Kansai).

Twelve fuel rods extracted from the lead fuel assembly at the reactor site were examined at the hot-cell facility of Japan Atomic Energy Research Institute (JAERI) in Tokai-mura.

Visually, the fuel rods appeared to be in good conditions, but some small spallings were observed at the second span from the top where oxide film was relatively thicker than other spans. Even in this span, the maximum oxide film thickness was less than 50 μm Fission gas release rate was less than 1%, which caused only a small increase in fuel rod internal pressure. Mechanical properties of the fuel cladding were evaluated by tensile tests.

These PIE results were within the range of other PIE data previously obtained from domestic and foreign PWR fuel rods. The data confirmed that the integrity of B-type fuel would be maintained at least up to 48 GWd/t.     

长寿命裂变产物核素核数据测量进展

文章对与高放废物深地层处置以及分离嬗变相关的半衰期大于10a、裂变产额高于0.01%的13种长寿命裂变产物核素的半衰期、裂变产额和热中子反应截面的测量研究、数据现状及其进展进行概要评述。就长寿命核素的分离纯化、原子数测定及放射性活度测量方法和技术进行了分析和论述。     

氢化锆用于小型钠冷快堆屏蔽的初步研究

为减少小型钠冷快堆（SSFR）堆侧的屏蔽厚度,本文选择氢化锆作为SSFR堆侧的屏蔽材料。使用一维离散纵标法（ANISN程序）计算了氢化锆在SSFR堆芯区能谱下的屏蔽特性,并计算了堆侧采用氢化锆和碳化硼的屏蔽厚度。结果表明：与堆侧采用碳化硼和不锈钢屏蔽相比,采用氢化锆和碳化硼屏蔽（碳化硼所占体积比小于0.3）,屏蔽厚度减小了大约20%。氢化锆和碳化硼混合屏蔽材料具有很好的屏蔽性能,可减小SSFR堆侧的屏蔽厚度。     

Diffusion and Evaporation of Fission Products in Coated Fuel Particles

Fractional releases of ¹³³Xe, ¹⁴⁰Ba and ⁸⁹Sr from slightly-irradiated pyrolytic-carbon-coated and SiC-coated particles were measured over a temperature range of 1,200°–1,750°C. The results are analyzed mathematically in order to obtain the diffusion and evaporation coefficients relevant to PyC and SiC. The resulting expressions for the coefficient of diffusion in PyC are 2.9x10-⁷ exp(-61x10³/RT) for ¹³³Xe and 4.7x10-² exp(51x10³/RT) for ¹⁴⁰Ba. For the coefficients of evaporation of ¹⁴⁰Ba from PyC, the expression is 3.5x10³ exp(-42x10³ /RT). As for SiC, the diffusion and evaporation coefficients of these nuclides are given for a temperature of 1,750°C. A high diffusivity path for the diffusion of ¹⁴⁰Ba is postulated to explain the difference in diffusion behavior between ¹³³Xe and ¹⁴⁰Ba in PyC.