Calculation of Beta-Ray Spectra from Individual and Aggregate Fission Products

The β-ray spectra of individual fission products were calculated by using the β-decay data assuming every β-decay to be allowed transition. For the nuclides without measured decay data the β-feeding function was evaluated with the gross theory of β-decay and the β-ray spectrum was calculated from the function. The measured decay data were also supplemented with the data calculated by the gross theory for the excitation energy range above the highest measured excitation energy level. The β-ray spectra from aggregate fission products after a burst fission were calculated by using the β-ray spectrum and the atom number of each fission product nuclide and they were compared with the ones measured for thermal neutron induced fission of ²³⁵U, ²³⁹Pu and ²⁴¹Pu at Oak Ridge National Laboratory. The spectrum calculations showed excellent agreement with the measured data at shorter cooling times than 10s when many short-lived nuclides without measured decay data contributed considerably to the spectrum.     

Reconsideration of the Theoretical Supplementation of Decay Data in Fission Product Decay Heat Summation Calculations

In the early stage of the development of the fission product (FP) decay data library for decay-heat summation calculations a serious disagreement between calculations and sample-irradiation measurements was experienced world wide. This problem was essentially circumvented by introduction of the nuclear model calculations. By applying the recent results using the total absorption gamma-ray spectrometer to the decay heat calculations we found that TAGS results seem to be free from the pandemonium problem and, in this respect, it provides a solid basis of the summation calculations without the supplementation of the nuclear model calculations. Among the typical FP nuclides which deserve the future TAGS experiment there are ⁹⁸Nb, ¹⁰⁰Nb, ¹⁰⁵Mo and ¹⁰²Tc and others from the present survey.     

Validation of decay heat calculation results of ORIGEN2.2 and CASMO5 for light water reactor fuel

Comparing with the fission product nuclide (FP) decay heat summation calculation result in MeV/sec/fission based on the JENDL FP decay and yield data files 2011 for the burst fission, FP decay heat calculated by ORIGEN2.2 coupled with JENDL-4.0 base library ORLIBJ40 was verified at the cooling time from 1 sec to 10⁸ sec for ²³⁵U (thermal), ²³⁸U (fast), ²³⁹Pu (thermal) and ²⁴¹Pu (thermal). For these fission nuclides, FP decay heat calculated by CASMO5 at the same cooling time after a short irradiation (10⁴ sec) was also compared with that of ORIGEN2.2. In the analysis of decay heat measurements at the cooling time from 2.3 years to 27 years consisting of four data sets on the fuel assemblies discharged from the US PWRs and BWRs, and the Swedish PWRs and BWRs, the average values of the ratios of the calculated to measured results (C/E's) were from 0.972 to 1.031 for ORIGEN2.2, and from 0.977 to 1.016 for CASMO5. The standard deviations of C/E's for the four data sets were from 0.02 to 0.03 for the both codes except for those of the US BWR fuel assemblies which were from 0.11 to 0.12. The obtained C/E's were similar to those in the precedent study.     

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions

For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 s or less of time lapse after an initial power burst.     

Uncertainty quantification of neutron multiplication factors of light water reactor fuels during depletion

Nuclear data-induced uncertainties of infinite neutron multiplication factors (k_∞) during fuel depletion are quantified in a single cell and a 3×3 multi-cell including burnable absorbers. Uncertainties of reaction cross sections, fission yields, decay half-lives and decay branching ratios provided in the JENDL libraries are taken into account. Hundred percent uncertainties are assumed to nuclear data to which uncertainty information are not provided in JENDL. Uncertainties propagation calculations are carried out with the adjoint-based procedure, and required sensitivity profiles of k_∞ with respect to these nuclear data are efficiently calculated by the depletion perturbation theory. Covariance matrices for fission yields and decay data in a simplified burnup chain are successfully generated by the stochastic-based procedure. k_∞ uncertainties of about 0.6% during fuel depletion are obtained, and it is shown that actinoids reaction cross sections are dominant contributors. Nuclide-wise decomposition of the uncertainties and observation of component-wise sensitivity profiles provide physical interpretations. By virtue of the adjoint-based procedure, several parametric surveys are also conducted. Contributions of uncertainties in fission products (FPs) nuclides are quantified, and important nuclides and energy ranges are identified for further evaluation of nuclear data of FP nuclides. Effect of cooling period on k_∞ uncertainties is also discussed.     

Uncertainty analyses of decay heat summation calculations using JENDL,JEFF, and ENDF files

The uncertainty analyses of decay heat calculation were carried out using major evaluated nuclear data files, JENDL, JEFF, and ENDF. The uncertainties were obtained from the sensitivity of individual fission product nuclide to the decay heat summation calculation. The summation calculation was performed for a burst fission. The sensitivities derived from the analyses were for decay energy, fission yield, and decay constant among the nuclear data included in the summation calculation. The uncertainties of the calculations at 0.1 s after a fission burst are ～10% for JENDL and ～8% for JEFF and ENDF and those at 10⁴ s are less than 2% for all cases. The main differences came from the different adoption of the energy uncertainty. The sensitivity analysis can be used to improve the decay data for decay heat calculation.     

Neutron energy dependence of delayed neutron yields and its assessments

Incident neutron energy dependence of delayed neutron yields of uranium and plutonium isotopes is investigated. A summation calculation of decay and fission yield data is employed, and the energy dependence of the latter part is considered in a phenomenological way. Our calculation systematically reproduces the energy dependence of delayed neutron yields by introducing an energy dependence of the most probable charge and the odd–even e?ect. The calculated fission yields are assessed by comparison with JENDL/FPY-2011, delayed neutron activities, and decay heats. Although the fission yields in this work are optimized to delayed neutron yields, the calculated decay heats are in good agreement with the experimental data. Comparison of the fission yields calculated in this work and JENDL/FPY-2011 gave an important insight for the evaluation of the next Japanese evaluated nuclear data library (JENDL) .     

Decay Heat Calculations Based on Theoretical Estimation of Average Beta- and Gamma-Energies Released from Short-Lived Fission Products

Fission product decay heat of ²³⁵U was calculated for short cooling-time on the basis of a preliminary version of a new decay data library recently completed by the Japanese Nuclear Data Committee. It was shown that a full adoption of recent publications of decay schemes to derive average energies of β- and γ-rays, ē_β, and ē_γ, leads to a large underestimation of the γ-ray component of the decay heat and to an overestimation of the β-ray component. In order to remedy this, theoretical values of ē_β and ē_γ were introduced for high Q-value decays, which were obtained with a gross theory of β-decay. It improved remarkably the agreement between calculation and experiment not only for the ²³⁵U decay heat but for ²³⁹Pu and ²⁴¹Pu as well. It was concluded that a large part of decay schemes recently published for high Q-value nuclides are inappropriate to use in calculations of ē_β and ē_γ, because they fail to reproduce the effect of β- strengths at high excitations, which makes ē_β small and ē_γ large. The use of the gross theory introduces this effect correctly into the values of ē_β and ē_γ and, hence, leads to a quite good prediction of both β- and γ-ray components of the decay heat.     

Calculation of neutron and gamma-ray energy spectra for fusion reactor shield design: Comparison with experiment-II

Measured and calculated neutron and gamma-ray energy spectra resulting from the transport of 14 MeV neutrons through a 0.30-m-thick lithium hydride slab and through a 0.05-m-thick lead slab followed by 0.30 m of lithium hydride are compared. Also reported are comparisons of the measured and calculated neutron energy spectra behind an 0.80-m-thick assembly comprised of stainless steel type 304 and borated polyethylene. The spatial dependence of the gamma-ray energy deposition rate measured using thermoluminescent detectors is compared with calculated data. The calculated data obtained using two-dimensional radiation transport methods and ENDF/B-IV cross-section data are in good agreement for all of the experimental configurations. Calculated integral neutron energy spectra agree with the measured data within 5 to 20% depending on neutron energy for the LiH and Pb plus LiH assemblies. The gamma-ray spectra agree within 20% for these slabs. The measured and calculated neutron energy spectra behind the SS-304-borated polyethylene assembly agree within 5% except at neutron energies below 5 MeV where background radiation influences the measured spectra. The gamma-ray energy deposition rates as a function of depth agree within a factor of two at all detector locations.     

基于衰变链计算的裂变产物缓发γ能谱分析

在军控核查技术中,缓发γ能谱是核材料的“指纹”。为计算和分析铀裂变产物的缓发γ能谱,本文将各种类型的衰变链简化为基态线性链和激发态线性链,推导了零时前后各级核素数目的变化公式,建构了计算缓发γ射线能谱的C语言程序代码,并通过实验对理论推导进行了验证。通过分析几种核素的缓发γ射线计数发现,计算结果与实验数据吻合较好。     

强中子场裂变产物中子核反应网络方程研究

本文建立了考虑中子参加反应的裂变产物中子反应及衰变的网络方程,选用求解一阶线性刚性微分方程组的Gear方法,开发了可计算任意裂变产物核数量在不同中子场强度和中子谱下随时间变化的核反应网络方程计算系统FIRENEQ,并配套了裂变产物产额和衰变数据库FPYDDL及裂变产物核中子反应截面数据库FPNCDL。检验结果表明,计算结果正确,程序可靠。利用该程序系统,研究了裂变产物核数量在不同中子场、不同诱发中子能量下随时间的变化。     

Evaluation of Neutron Capture Gamma-ray Spectra for 89Y, 93Nb, 127I, 133Cs, 141Pr, 197Au,natTl, and 209Bi

The neutron capture gamma-ray spectra for 8 nuclides, ⁸⁹Y, ⁹³Nb, ¹²⁷I, ¹³³Cs, ¹⁴¹Pr, ¹⁹⁷Au, ^natTl, and ²⁰⁹Bi, were calculated by using the Hauser-Feshbach statistical model, and their results were compared with the available experimental data. Two dominant ingredients to perform the statistical calculation were the level densities described by the Gilbert-Cameron approach with an improved systematics, and the gamma-ray transmission coefficients described by gamma-ray strength functions. Although various gamma-ray strength functions with a Lorentzian formula have been developed by using the photonuclear data or a microscopic analysis, they have failed to reproduce the occasional anomalous bumps observed near or below a neutron binding. In this work, we could reproduce the bumps well by adding a Lorentzian with an energy-temperature dependent width into a giant electric dipole resonance with an enhanced generalized Lorentzian. In addition, we introduced a correction function so as to compensate for the shortcomings due to missing levels or level-cuts.     

Evaluation of Shutdown Gamma-ray Dose Rates around the Duct Penetration by Three-Dimensional Monte Carlo Decay Gamma-ray Transport Calculation with Variance Reduction Method

For the evaluation of gamma-ray dose rates around the duct penetrations after shutdown of nuclear fusion reactor, the calculation method is proposed with an application of the Monte Carlo neutron and decay gamma-ray transport calculation. For the radioisotope production rates during operation, the Monte Carlo calculation is conducted by the modification of the nuclear data library replacing a prompt gamma-ray spectrum with a decay gamma-ray spectrum. By multiplying each correction factor, which is ratio of the actual activation level after shutdown to the production rate during operation, with each decay gamma-ray flux due to each radioisotope, the decay gamma-ray dose rate is evaluated. In order to improve the statistical error, a variance reduction method is proposed by the application of the weight window importance technique and the specification of the decay gamma-ray generation location. We identify the cell producing the decay gamma-ray which can contribute the decay gamma-ray flux in evaluation locations, and forcibly terminate the gamma-ray transport calculation in the cells except for the identified cells. In order to validate the effectiveness of the method, shielding calculation for actual ITER (International Thermonuclear Experimental Reactor) configuration is performed, and small statistical errors below criteria are obtained. The effectiveness of the proposed method for ITER design analysis is demonstrated.     

一个利用DEnet计算机网络的在线核衰变能谱自动测量程序

本文介绍的应用程序是将基于PDP11/34计算机数据获取系统的在束多参数事件测量变换为离束在线-序列核衰变谱的自动测量。如此测量的数据立即通过DECnet计算机网络转储到VAX11/780计算机中,以深受物理工作者欢迎的数据文件形式存贮在用户的帐号之下。     

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.     

通用级联衰变动力学简化计算

级联衰变动力学方程是描述放射性同位素衰变规律的基本方程之一,衰变过程的复杂性和衰变子体的多样性,使级联衰变动力学计算十分复杂。本工作分3种情况给出通用级联衰变动力学解析解表达式,用矩阵函数对其进行形式简化,使级联衰变动力学计算简单明了,便于应用;对级联衰变公式进行简单讨论,并利用该公式求解了²³⁹Pu级联衰变中各级子体的原子数及²³⁹Pu的衰变热功率。     

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory-scale reprocessing, the release ratio was measured for fission product (FP) nuclides such as Ru, Tc-99, Cs, Sr, Nd, Y, Mo, Rh and actinides such as Cm-242 and Am-241. As a result, the release ratio was 0.20 for Ru and was around 1×10^?4 for the FP and actinide nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to its initial concentration in the test solution. The release ratio decreased with the increase in the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the increase in the initial concentration. The release ratio of Ru and NOx concentration increased with the increase in the temperature of the test solutions. They were released together almost at the same temperature between 200 and 300 °C. Size distribution of particles like mist was measured. The data show that there was a difference between distributions at the temperatures below 150 °C and over 200 °C.     

Nuclear data sheets for A = 133

Nuclear structure data are compiled for the ten known members of the A = 133 mass chain. Decay and reaction data from approximately 200 references available through July 1973 are evaluated. Adopted level properties and gamma-ray properties are presented in summary form for each nucleus. Included are seven drawings which show most of the decay properties and adopted values. Following the data section for each radioactive nucleus is a discussion of its decay scheme. Since data for short-lived members of the A-chain are limited and sometimes inconsistent, few definitive spin and parity assignments can be made for their daughter nuclei.     

Improvement of photonuclear reaction model below 140 MeV in the PHITS code

The photonuclear reaction model in the particle and heavy ion transport code system (PHITS) code is improved for incident photon energies below 140 MeV. Japanese Evaluated Nuclear Data Library (JENDL) Photonuclear Data File 2004 (JENDL/PD-2004) is adopted to determine the total reaction cross section. The statistical decay model after excitation of the nucleus in PHITS is improved by modifying the decay widths for light nuclei under the isospin selection rule to reasonably reproduce the proton and neutron emission in the de-excitation process. The quasideuteron disintegration process is newly introduced into PHITS to handle the photonuclear reaction up to 140 MeV of incident photon energy. The accuracy of the improvements was verified by comparison with the experimental literature data. The improved PHITS can contribute to various practical applications such as neutron dose estimation in X-ray therapy.     

Calculation of low-energy electron antineutrino spectra emitted from nuclear reactors with consideration of fuel burn-up

Nuclear reactors produce a great number of electron antineutrinos mainly from beta-decay chains of fission products (FP). Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may have information on fuel burn-up and may be detected in the future with advanced measurement technology. We calculate reactor antineutrino spectra especially in the low-energy region. In this work, we present neutrino spectra from various reactors such as typical pressurized water reactor and others types of reactors for comparison. Our result shows the electron antineutrino flux in the low-energy region increases with burn-up of nuclear fuel by accumulated nuclides with low Q values in beta-decay.