Radiation damage and simulated fission product effects on the properties of inert matrix materials

Candidate inert matrix materials for actinide transmutation (MgAl₂O₄, CeO₂) or immobilization (ZrSiO₄) containing ²⁴¹Am were characterized. The currently most considered material, ZrO₂, was produced, with La₂O₃ as stand-in for Am, and with and without simulated fission products to investigate burnup effects. The oxygen potential was measured using an EMF cell. The accumulation of radiation damage due to Am decay was investigated by periodically measuring lattice parameters and hardness. Sequential leaching tests in deionized water, aimed at correlating the leaching behaviour of Am and of the matrix with radiation damage, showed significant release of Am and of some matrix components.     

Influence of evaluated data of fission product poisons on criticality

The main objective of this research is to study the influence of cross-section differences of fission product poisons among various newly released evaluated cross-section libraries ENDFB-VI.8, JENDL3.2, JEF2.2, IAEA, ENDFB-VII and JEFF3.1 on criticality of an MTR type research reactor. The effect of cross-sections of poisons on the reactivity was studied with the help of WIMSD and CITATION codes. Various cross-section libraries were used in SARC (System for Analysis of Reactor Core) code for the fuel cycle analysis. It was found that the negative reactivity induced due to ¹³⁵Xe for the equilibrium core is around 36.00 mk whereas for ¹⁴⁹Sm it ranges from 6.65 to 7.06 mk. The three libraries (JENDL3.2, IAEA and ENDFB-VII) resulted in small increase in the Xenon worth as compared to the other three libraries. For Samarium, JEFF3.1 gives the highest worth whereas ENDFB-VI.8 gives the least worth among the six libraries.     

Influence of corium oxidation on fission product release from molten pool

Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.     

~(147)Pm对小鼠生殖细胞的遗传毒理效应

本文研究观察了小鼠静脉摄入重核裂变产物~(147)Pm后不同间隔时间对生殖细胞染色体畸变和精子(主要是无钩精子)畸形的诱变效应,实验结果发现~(147)Pm可诱发精原细胞染色体结构畸变;随着机体受内污染时间的延长,染色体畸变率和多倍体细胞也有所增加,同时~(147)Pm也可诱发初级精母细胞产生染色体断片和易位,形成多价体。染色体断片率随~(147)Pm辐照时间的延长而升高。而多价体只在内污染10天后的实验组中出现。     

Volatile fission product and sodium release from liquids

For the improvement of radioactive source term calculations the computer code revols has been developed for the mechanistic modeling of the evaporative release of volatible species (e.g. water, sodium and volatile fission products as NaJ, Cs and Rb) from different hosts into an inert gas atmosphere. The code, showing a modular structure, has been developed to be coupled with reactor containment safety analysis codes as the contain / lmr and lmfbr version. In substituting existing constant-retention-factor formulations by introducing a geometry and state dependent, instationary retention factor, an improved aerosol and fission product source calculation can be obtained. The comparison of theoretical predictions with experimental results performed at the Karlsruhe Research Center shows good agreement.     

Diffusion behavior of fission product in pyrolytic silicon carbide

The concentration distributions of ¹³³Xe, ¹⁴⁰Ba, ⁸⁹Sr, ¹⁴¹Ce, ¹⁰³Ru and ⁹⁵Zr-⁹⁵Nb in a SiC layer and releases of these fission products from SiC coated fuel particles were measured in a temperature range from 1650° to 1850°C, to obtain the diffusion parameters and to investigate the diffusion behaviors in SiC. Temperature dependences of the diffusion coefficients of these fission products except ⁹⁵Zr-⁹⁵Nb were obtained. In diffusion behaviors of the alkali earth fission product, volume diffusion occured considerably besides grain boundary above 1650°C. ¹⁴¹Ce and ¹⁰³Ru diffusions were almost through grain boundaries in the temperature range. ⁹⁵Zr-⁹⁵Nb showed volume diffusion besides boundaries diffusion in their distribution annealed at 1850°C for 35 h, but those for the shorter time or at the lower temperature would be due to boundary diffusion. In comparison of the diffusion coefficients in SiC and PyC, it was proved that SiC was effective for retention of the solid fission products.     

Experimental study on the effect of fission product gas release into blockage wake region using a simulated LMFBR fuel subassembly

The objective of this study is to evaluate temperature rise due to gas release in the wake region of LMFBR fuel subassemblies. The experiments were conducted in two sets of grid-spacer-type 37-pin bundles simulating LMFBR fuel subassemblies. In Test section 37GC, the central 24 subchannels were blocked by a stainless steel plate and in Test section 37GE one-half edge part (39 subchannels) of the total flow area was blocked by the same material. The experimental results were compared with data obtained in similar tests using a spacer wire-type pin bundle, designated 37WC. The temperature rises in 37GE and 37WC were nearly identical in value and effect of gas release rate. The marked agreement seems to imply that there is a limit in the content of released gas in the wake region. On the other hand, the temperature rise behind the central blockage in the grid-type bundle, where gas might easily flow out to the core flow region, was far less than in the other geometries.     

Radiation re-solution of fission gas in uranium dioxide and carbide

Calculations have been performed to estimate the removal rate of fission gas atoms from bubbles due to collisions with energetic fission fragments and recoil cascades. The efficiency of this process was found to be higher than estimated earlier, but is still too low to be responsible for the experimental observations of fission gas bubble destruction during irradiation of oxide fuel. An irradiation experiment to investigate the interaction of fission spikes with free surfaces has enabled a simple theory to be developed which can explain the shrinkage of bubbles and pores by the surface relaxation of a shock wave produced by the passage of a fission fragment. This mechanism occurs in oxides but not carbides because of the faster dispersion of the fission fragment energy and provides the major reason for the difference in gas bubble distributions in oxide and carbide fuel. This process, however, does not remove gas atoms from the bubbles. Since high levels of apparently diffusive fission gas release are observed in oxides, the “effective solubility” of the fission gases required for this release must be sought in phenomena other than the fission spike.     

14.8 MeV中子诱发~(238)U裂变产物产额测量

裂变产物产额作为裂变过程的一个重要参数,其准确测量对有关裂变的很多方面都有重要意义。为了准确测量中子诱发~(238)U裂变产物产额,利用中国工程物理研究院PD-300加速器上的T(d,n)4He反应,产生14.8MeV的中子,诱发~(238)U裂变。辐照过程中,通过金硅面垒半导体探测器监测中子通量的变化。使用Al片作为监测片计算整个照射过程中样品的平均中子通量。辐照结束后,利用高纯锗(High-Purity Germanium,HPGe)探测器测得裂变产物特征γ射线计数,计算得到裂变产物的产额,使用MCNPX软件对中子的多次散射和自屏蔽效应进行修正,并通过计算得到样品和监测片的自吸收修正、中子通量波动因子。得到了95Zr、127Sb、140Ba、147Nd、131I、103Ru等长半衰期产物的累积产额值,并将结果与以前的文献值做了比对,研究结果有助于~(238)U裂变产物产额的分析和评价。     

保偏光纤电离辐射效应研究

本文从光纤陀螺中主要部件保偏光纤空间应用的角度出发,利用60Co辐照源模拟空间辐照,对保偏光纤环的电离辐射效应展开研究,基于保偏光纤在不同剂量率条件下的辐照试验,探索保偏光纤随辐照剂量的变化规律。实验结果表明,累积剂量和剂量率为63Gy与0.01Gy/s、32Gy与0.02Gy/s、326Gy与0.0835Gy/s的光纤损耗分别增加1.0dB、0.55dB和5.8dB,也即保偏光纤损耗随总剂量与剂量率增大。理论分析得到的光纤辐照损耗与剂量率的关系与试验数据相吻合。本文结果可为保偏光纤电离辐射效应的地面试验提供参考。     

The effect of irradiation-induced re-solution on fission gas release

A finite-difference technique is used to compute exact solutions to the diffusion equation describing fission gas release from UO₂ nuclear fuel during steady reactor operation. The resolution of gas atoms from grain-boundary bubbles is treated in two alternative ways, and the results of the parallel calculations compared. Predictions of gas release using simple analytical models are compared with the numerical results and are found in general to describe the process very accurately.     

Development and application of the PBMR fission product release calculation model

At PBMR, long-lived fission product release from spherical fuel spheres is calculated using the German legacy software product GETTER. GETTER is a good tool when performing calculations for fuel spheres under controlled operating conditions, including irradiation tests and post-irradiation heat-up experiments. It has proved itself as a versatile reactor analysis tool, but is rather cumbersome when used for accident and sensitivity analysis. Developments in depressurized loss of forced cooling (DLOFC) accident analysis using GETTER led to the creation of FIssion Product RElease under accident (X) conditions (FIPREX), and later FIPREX-GETTER. FIPREX-GETTER is designed as a wrapper around GETTER so that calculations can be carried out for large numbers of fuel spheres with design and operating parameters that can be stochastically varied. This allows full Monte Carlo sensitivity analyses to be performed for representative cores containing many fuel spheres. The development process and application of FIPREX-GETTER in reactor analysis at PBMR is explained and the requirements for future developments of the code are discussed. Results are presented for a sample PBMR core design under normal operating conditions as well as a suite of design-base accident events, illustrating the functionality of FIPREX-GETTER. Monte Carlo sensitivity analysis principles are explained and presented for each calculation type. The plan and current status of verification and validation (V&V) is described. This is an important and necessary process for all software and calculation model development at PBMR.     

Modeling of fission product release from HTR fuel for risk analyses

In the United States and Federal Republic of Germany, methodologies have been developed to determine the performance of and fission product release from TRISO-coated fuel particles under postulated accident conditions. The paper presents a qualitative and quantitative comparison of models used by General Atomics (GA) and by the German Nuclear Research Center at Jülich (KFA/ISF). A benchmark calculation was performed for fuel temperatures predicted for the U.S. Department of Energy (DOE) sponsored Modular High-Temperature Gas-Cooled Reactor (MHTGR). Good agreement in the benchmark calculations supports the on-going efforts to verify and validate the independently developed codes of GA and KFA/ISF.