SEAFP Modeling of Accident Sequences, Source Terms, and Public Doses

Within the European Safety and Environmental Assessment of Fusion Power (SEAFP), off-site public doses were assessed for representative hypothetical worst case fusion power station accident sequences driven by in-plant energies, without taking credit for any active safety measures. In this paper, in order to illustrate the calculations performed in SEAFP, the calculational sequence is described for one accident scenario. This is a major in-vessel LOCA. Several sources of active material are mobilized following the LOCA, and are transported across successive containment barriers as the accident evolves, with a small fraction of the source inventory eventually reaching the environment. Using conservative assumptions, modeling of thermo-fluid-mechanics, heat transfer, mobilization, transport, aerosol phenomena, and atmospheric dispersal and dilution, were used to determine several measures of public dose exposure. Calculations for other accident scenarios, performed within SEAFP, are not described in detail in this paper, but are commented on. The calculations indicate that maximum public doses would be well below levels at which emergency intervention would be required.     

Source Term and Mobilization Assessment in NSSR-1 of ITER

We describe the radioactive sources in the International Thermonuclear Experimental Reactor (ITER). The most important sources are co-deposited tritium, tritiated water, tokamak dust, and corrosion products. The co-deposited tritium is limited to 1 kg-T; the total on-site tritium inventory in the Basic Performance Phase (BPP) is 4 kg-T. Tritiated water concentrations are kept below 0.2 g-T/m³ in the divertor; other coolant loops have lower tritium concentrations. The in-vessel dust inventory is up to 100 kg-W, 100 kg-Be, and 200 kg-C. The activated corrosion product inventory is kept below 10 kg per loop.     

水冷聚变堆主回路活化产物源项计算分析

水冷聚变堆中结构材料活化腐蚀产物和冷却剂活化产物是正常运行工况下的最主要放射性来源,也是反应堆运行及维护过程中工作人员辐照剂量的直接来源。本文使用CATE V2.1程序对国际热核聚变实验堆(International Thermonuclear Experimental Reactor,ITER)LIM-OBB(Limiter-Out-Board Baffle)冷却回路的活化腐蚀产物和水活化产物进行模拟计算,并根据CATE模拟得到的放射性活度通过点核积分程序分别计算正常运行1.2 a及停堆15 d的剂量率。计算结果表明,反应堆运行期间冷却剂活化产物比活度和剂量率远大于结构材料活化腐蚀产物,而停堆后冷却剂活化产物迅速衰变完,结构材料的活化腐蚀产物成为比活度和剂量率的主要来源。     

Uncertainty and range of alternatives in estimating tritium emissions from proposed fusion power reactors and their radiological impact

A review of technology assessment studies as well as original research papers is presented in two diagrams with respect to estimating (1) the performance of fusion reactors in containing tritium, and (2) the radiological consequences of normal operational losses as well as accidental releases of tritium. Predictions are shown to vary over several orders of magnitude.This paper is based on a more comprehensive study.²³     

Overview of Safety and Environmental Issues for Inertial Fusion Energy

This paper summarizes safety and environmental issues of Inertial Fusion Energy (IFE): inventories, effluents, maintenance, accident safety, waste management, and recycling. The fusion confinement approach among inertial and magnetic options affects how the fusion reaction is maintained and which materials surround the reaction chamber. The target fill technology has a major impact on the target factory tritium inventory. IFE fusion reaction chambers usually employ some means to protect the first structural wall from fusion pulses. This protective fluid or granular bed also moderates and absorbs most neutrons before they reach the first structural wall. Although the protective fluid activates, most candidate fluids have low activation hazard. Hands-on maintenance seems practical for the driver, target factory, and secondary coolant systems; remote maintenance is likely required for the reaction chamber, primary coolant, and vacuum exhaust cleanup systems. The driver and fuel target facility are well separated from the main reaction chamber.     

中子活化分析在大气环境及水环境研究中的应用

利用中子活化分析方法测试了成都市的大气环境及水环境样品。研究了有害污染元素分析的特征。     

Fusion safety and environmental R&D tasks in Japan

This paper aims at listing and evaluating the status of all the research and development (R&D) tasks necessary for the construction of a safe and environmentally benign fusion experimental reactor. At this time, it is not possible to define precisely the R&D tasks necessary for the licensing approval and those that are useful in improving safety but not necessarily required for licensing because the licensing procedure itself is still being discussed. Among the R&D tasks, the most important are considered to be those related to tritium safety, namely, those effective in reducing the uncertainty in tritium inventory in the plasma facing components and blanket, uncertainty in tritium permeation and leakage, and those to clarify tritium behavior in the containment and in the environment. The R&D tasks with second priority are judged to be those related to mobilization of the activation products such as activated erosion dust or the corrosion products. The volatilization of structural metal caused by the oxidation at high temperature seems to be highly unlikely but some experiments are needed to assure that this is the case.     

Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.     

超热中子活化分析用于燃煤电厂对环境土壤污染的研究

大燃煤电厂每天由烟囱排出的粉尘(飞灰)有几十吨,通过干沉降与湿沉降,它们最终将落在周围土壤表面上。由于地质构成上的不同,一般土壤中的元素浓度与煤飞灰中的元素浓度是不同的,大量飞灰的下落,以及雨水的冲刷、渗透,可能会改变土壤中原有元素的浓度,从而影响人体健康。这是人们关心的问题。     

液闪法测氚的一些优化条件和质量检验研究

对液闪法测氚的一些影响因素进行了优化条件实验，通过比对及内部检验证明在优化的实验条件下液闪法测氚所得分析结果的平行性、重现性很好。所用标准源成分如与试样不同，将带来系统误差，应予修正。     

An evaluation of the influence of impurity concentration and composition in concrete on activation and radiation dosein the reactor facility

The content analysis of radioactive waste and radiation dose evaluation is considered as one of the important factors in the reactor facility design.This kind of buildings consists of the concrete for the most part and uses it as the structure and shield of the building.Generally,the concrete has impurities such as cobalt,europium,nickel,and cesium with specific content depending on the production method or manufacturing company.Dominant radioactive nuclides generated from the fundamental compon...     

The relative importance of collective and individual doses from routine releases of radioactive materials to the atmosphere

Two radiological quantities, the dose to the critical group and the collective dose to the exposed population, are of interest in assessing the significance of discharges of radionuclides to the atmosphere. This paper examines two aspects of these quantities, the distance over which collective dose must be integrated to ensure that the integral has converged and the extent to which collective dose is effectively limited because of regulatory control on the dose to critical groups.

For most nuclides an integration to 1000 km is shown to be adequate when assessing collective dose. Regulatory control of airborne discharges has generally limited the critical group doses to small fractions of the ICRP recommended dose limit, when the resulting collective dose is unlikely to exceed a few man-Sv. Given the costs typically involved in waste management there can be little scope for the further cost-effective reduction of airborne discharges controlled on the basis of such doses to the critical group.     

Measurement and calculations of long-lived radionuclide activity forming in the fast neutron field in some ITER construction steels

Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, for some ITER construction steels. The activation was conducted in fast neutron irradiation channel of the MARIA research fission reactor (Poland). The dimensions of steel samples were 10 mm × 10 mm × 1 mm and mass was approximately 0.8 g. The neutron flux density was measured by means of activation foil method and unfolding technique; fraction of neutrons above 1 keV was 95%. The activation lasted 242 h and cooling took 100 days; the mean neutron flux density was 2.9E12 n/(cm² s) (neutrons above 500 keV are 53% of total) whereas total fluency 2.53E18 cm⁻². The activity measurements were done by means of gamma-ray spectrometry. Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TENDL-2011 and experimentally determined neutron flux. Measured activity of long-lived gamma emitting radionuclides was, in average, about 6.3 MBq/g 100 days after activation; the dominant radionuclides were ⁵⁸Co and ⁵⁴Mn (about 81% and 14% of total activity respectively). The C/E ratio differs for particular radionuclides and is in the range 0.86–0.92 for ⁵¹Cr, 0.93–1.21 for ⁵⁴Mn, 0.77–0.98 for ⁵⁷Co, 0.91–1.21 for ⁵⁸Co, 1.17–1.27 for ⁵⁹Fe, and 1.75–2.44 for ⁶⁰Co.     

通古斯卡爆炸现场泥炭样品的中子活化研究

用中子活化法研究了取自１９０８年通古斯卡爆炸现场一个钻孔泥炭样品中的铱和其它有关元素的丰度变化,发现在１９０８年爆炸层及其上覆层位处的铱有明显的异常,最大铱丰度约为背景值的５倍以上。     

Measurement of trace element content in bone samples using long-lived neutron activation products and high-resolution gamma-ray spectrometry

A nondestructive instrumental neutron activation analysis with high-resolution gamma-ray spectrometry of long-lived radionuclides was developed and used for measurement of trace element contents in samples of bone to determine health and diseases. Using this method, the silver(Ag), cobalt(Co), chromium(Cr), iron(Fe), mercury(Hg), rubidium(Rb), antimony(Sb), selenium(Se), and zinc(Zn) mass fractions were estimated in bone samples from 27 patients with intact bone(12 females and 15 males, aged from 16 to 49 years) who had died from various non-bone-related causes, mainly unexpected traumas,and from 5 patients with chondroma(2 females and 3males, 15–42 years old), obtained from open biopsies or after operation. The reliability of the differences in the results between intact bone and bone affected by chondroma was evaluated by a parametric Student's t test and a nonparametric Mann–Whitney U test. It was found that in the bone affected by chondroma, the mean mass fractions of Co, Cr, Fe, Se, Sb, and Zn were significantly higher than in normal bone tissues. In the neoplastic bone, many correlations between trace elements found in the control group were no longer evident. This work revealed that there is asignificant disturbance of the trace element metabolism in bone affected by chondroma.     

核动力堆安全壳内外中子能谱和剂量测量

利用自制的多球谱仪测量了某核动力反应堆安全壳内外的中子能谱和剂量当量率。对安全壳外测量，中心探测器为球形^3He正比计数管；对安全壳内测量，中心探测器为球形金箔。系统的响应函数用MCNP程序计算，解谱程序为MIEKEB。为验证系统响应函数计算的准确性，进行了一些实验测量，并与理论计算结果进行了比较。结果表明，测量结果与计算结果在不确定度范围内相吻合。     

Estimation of infiltration and surface run-off characteristics of radionuclides from gamma dose rate change after rain

A method to estimate the infiltration and surface run-off characteristics of radionuclides on three types of ground surface from gamma dose rate change due to rain has been developed. We proposed the estimation methods based on the differences in the dose rate increases between monitoring stations caused by different attenuation of gamma ray due to infiltration and by the different run-off characteristics. The gamma dose rate data used for the estimation were measured at near-by monitoring stations in which the ground type around the detector differed from each other. Rain events were selected by the criteria in which ²¹⁴Pb and ²¹⁴Bi deposition amounts are considered to be uniform with in the area in which objective monitoring stations are deployed. We also calculated the surface concentration considering both infiltration and surface run-off processes. As a result, it was shown that the calculated surface concentration was about 40% larger than that considering neither processes of the infiltration and surface run-off.     

Neutron shielding and activation of the MASTU device and surrounds

A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day.The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.     

Transport of hydrogen and deuterium in the reduced activation martensitic steel ARAA

Advanced reduced activation alloy (ARAA) is a reduced activation ferritic/martensitic (RAFM) steel under development at the Korea Atomic Energy Research Institute. The transport of hydrogen and deuterium in ARAA was investigated in an elevated temperature range of 250–600 °C. A continuous-flow method, a time-dependent gas-phase technique, was used for the measurements. Complete sets of transport parameters (permeability, diffusivity, solubility, trap site density, and trapping energy) of hydrogen and deuterium in ARAA were successfully obtained. We show that appreciable trapping effects are observed only at low temperatures (250–350 °C) and that the isotope effect ratio for the diffusivity differs from the classical prediction. However, the measured values of permeability, effective diffusivity, and effective solubility of ARRA were within the range of results reported for other RAFM steels.     

Axial and radial distribution of thermal and epithermal neutron fluxes in irradiation channels of the Ghana Research Reactor-1 using foil activation analysis

The experimental method (foil activation) was used to determine the neutron fluxes in two outer irradiation channels of the Ghana Research Reactor-1. In the experimental procedure, it was observed that the fluxes rise to a peak before falling and then finally leveling out, axially. Axially and radially, it was also observed that the fluxes in the center of the channels were lower than those on the sides. Radially, the fluxes dipped in the center while they increased monotonically towards the sides of the channels. The results have shown that there are flux variations within the irradiation channels at both axial and radial directions.