ORE Assessment Due to ACP in the PHTS of the Point Design Phase of the ITER Project

The work presented here dealt with the revision and the updating of the ORE (Occupational Radiation Exposure) assessment for the ITER PHTS (Primary Heat Transfer System). The data used come from the Point Design Documents and refers to the ITER design of the first half of 1996. The MCNP computer code was adopted to perform the shielding calculation. In addition, an accurate approach to evaluate the photon flux during maintenance and inspection activities was followed and recently published photon-flux-to-dose-rate conversion factors were applied to obtain the corresponding dose rate. The ACP inventory was taken from the relevant calculation performed with the PACTOLE code for the Point Design. A special ACP calculation was performed for each PHTS circuit and the related results are used in the respective dose rate calculations. The collective dose for the main activities performed to maintain the PHTS components is reported. The dose result for each activity type is shown and the comparison with a reference fission plant is discussed.     

压水堆主冷却剂中氚源项计算分析

压水堆主回路冷却剂流经堆芯时,水中固有及特加核素受中子辐照后会产生氚,氚几乎全部以气体和液体的形式排入环境,造成氚污染。因此,氚是压水堆辐射环境影响评价的主要关注内容之一。本文以AP1000为例,根据压水堆主回路冷却剂中氚的产生途径及其随时间的变化情况建立详细的计算模型,计算压水堆主回路冷却剂中的氚活度并分析各产氚途径对氚产生量的贡献。计算结果表明：主回路冷却剂中的氚主要来源于可溶性硼的中子活化和铀裂变,对氚产生量的贡献达80%以上;在⁷Li纯度为99.9%时,AP1000主回路中的年产氚量为5.23×10¹³ Bq,锂产氚量占总量的14.01%,随⁷Li纯度的增加,锂产氚量的贡献呈线性减小,在⁷Li纯度为99.99%时,锂产氚量占总量的3.18%。其他途径对氚的产生量贡献很小,可忽略。根据以上结果,可通过控制主回路冷却剂中添加的初始硼浓度、提高燃料包壳质量、增加LiOH中⁷Li的纯度等多种途径来降低主冷却剂中氚的产生量,从而减少氚对环境的放射性污染。     

压水堆3H源项计算分析

对压水堆中氚的产生和消减机理进行了研究。根据一回路冷却剂中氚的代谢机制建立氚计算模型,分析了压水堆各途径对氚的产生量贡献及⁷Li纯度对锂产氚量的影响。结果表明：计算模型详细考虑了产生氚的核素随时间的衰减变化,计算的氚产生量为52.08 TBq/a。压水堆一回路冷却剂中的氚主要来源于可溶硼的中子活化反应、铀核的三元裂变,对氚产生量的贡献达90%以上,⁷Li纯度为99.9%时锂产氚量占总量的7.45%,其他途径对氚的产生量贡献很小,可忽略。锂产氚量的贡献随着⁷Li纯度的升高而线性减小。研究结果可为压水堆氚源项的计算提供参考。     

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.D.T.I. Dr. Trippe Ing. GmbH, Karlsruhe.     

Using the tritium plasma experiment to evaluate ITER PFC safety

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 10²³ ions/m².s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.Prepared for the U.S. Department of Energy, Office of Energy Research under DOE Idaho Field Office Contract DE-AC07-76ID01570.     

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.     

压水堆核电站一回路活化腐蚀产物源项控制措施探讨

材料替代和一回路水化学控制是降低活化腐蚀产物源项的主要措施。本文介绍M310、AP1000和EPR三种压水堆核电站一回路水化学优化情况,比较三种压水堆一回路活化腐蚀产物源项,分析探讨水化学优化对源项降低的影响,最后对国内压水堆核电站一回路水化学优化提出建议。     

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.     

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。     

ITER中国液态锂铅实验包层模块活化特性分析与废料处理

使用中子学程序系统VisualBUS和活化数据库EAF-99对DFLL-TBM的高级子模块DLL-TBM的活化特性进行了计算和分析,包括DLL-TBM各部件在不同停堆时间的活度、衰变余热和剂量率.活化计算所需要的三维中子能谱通过MCNP/4C中子/光子输运程序和国际原子能机构发布的FEND1.0数据库计算得到.在活化计算分析的基础上,参照欧洲聚变堆安全和环境评估(SEAFP)策略中有关核废料的处理标准评估了TBM各区材料在退役后的废料处理工作,包括核废料应该采用何种适当的方式进行处理及其被完全清除干净的可行性.     

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.     

Preliminary study on capsule material for ITER neutron activation system

A neutron activation system (NAS) measures neutron fluence at the first wall and the total neutron flux from plasma, providing the fusion power evaluation. A pneumatic transfer method is conventionally utilized to transfer encapsulated activation samples between the irradiation stations and counting station. The temperature of the irradiation station, near the first wall could reach too high for the conventional polymer-based materials, such as polyethylene, to be used as a capsule material for the ITER NAS. Considering the environment of the irradiation station of the ITER NAS, the candidate capsule materials are chosen as four materials: RAFM (reduced activation ferritic martensitic) steel, SiC_f/SiC composite, tungsten, and CFC (carbon fiber-reinforced carbon). Preliminary investigation reveals that the CFC is the most promising capsule material for ITER NAS due to its good thermal and magnetic properties as well as low activation by neutron irradiation. Various kinds of mock-up capsules are fabricated using CFC with the consideration of the volume of inner space accommodating activation samples. Preliminary pneumatic transfer experiments carried out in the small-scale test-bed suggest that the transfer speed of capsule should be slower than 10 m/s and the wall thickness of the capsule should be thicker than 2 mm so as not to be broken by impact damage. The present study shows the feasibility of using CFC as a capsule material for the ITER NAS.     

Status and verification strategy for ITER neutronics

The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6 MeV) gamma-rays are recommended.     

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.     

Lessons Drawn from ITER and Other Fusion International Collaborations

The international character of fusion research and development is described, with special emphasis on the ITER (International Thermonuclear Experimental Reactor) joint venture. The history of the ITER collaboration is traced. Lessons drawn that may prove useful for future ventures are presented.     

低活化马氏体钢在ITER水环境中的腐蚀性能研究

低活化马氏体钢作为聚变堆候选的结构材料,其腐蚀性能影响冷却回路辐射场的分布。本文选取CLAM、CNS-1和SCRAM-9 3种钢材,并用T91与3种钢材进行对比,分析4种材料的性能。整个实验回路温度维持在150 ℃,溶氧量小于0.01 mg/kg,pH值为7（20 ℃）,电导率小于1 μS/cm,压力为1 MPa,水的流速设为6 m/s。实验后,所有样品均采用失重分析法、XRD、EDS及SEM分析。结果显示,随着时间的增加,材料的失重量增加,腐蚀速率减少。4种材料的失重量均遵循幂函数规律,T91钢的耐腐蚀性较另3种钢好,而3种材料中CLAM的抗腐蚀性能相对更好;样品表面氧化层变得越来越致密且非常薄。氧化层腐蚀产物主要是Fe₂O₃和Fe₃O₄。     

Overview and status of ITER internal components

The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R&D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.     

Assessment of ITER PF coil quality from magnetic measurements

A feasibility has been demonstrated for numerical reconstruction on the base of magnetic measurements for geometrical displacements or deformations occurred in the manufacture and assembly of magnet coils. For validation of the proposed approach the test results of reconstruction of possible misalignments and deviations of the ITER PF1 coil are presented.     

ITER的靶丸加料计算(英文)

运用考虑动力效应的Kuteev2-D透镜模型,数值计算了靶丸在国际热核实验堆(ITER)中的消融率,讨论了目前现有的加料工艺的技术困难和可能的解决办法。数值积分结果发现目前已有的靶丸加料技术很难满足堆级等离子体ITER中心加料的要求,计算表明对一个2m长的单级气动枪要加速一个半径0.5cm的靶丸达到速度24.27km/s才能渗透ITER等离子体100cm。用两种典型的消融理论计算了渗透深度与靶丸速度和半径的依赖关系并作了比较。新近的研究从高场侧(HFS)注入靶丸来改善芯部加料效率可能给芯部加料困难贡献一种解决办法,对相关的问题作了讨论。     

Study of capsule position monitoring methods for ITER neutron activation system

The reliable monitoring of the position of an encapsulated activation sample is essential to ensure the diagnostic accuracy and the maintenance of the ITER neutron activation system (NAS). Conventional methods using optical or electrical detectors to determine the capsule position is difficult to be used in the ITER NAS because of limited space as well as extremely high electromagnetic and radiation environment. In this study, new methods using the flow rate change inside a transfer tube assembly and the propagation characteristics of sound wave are investigated for the reliable determination of the capsule position. Experimental results confirm that the abrupt reduction of flow rate in the transfer tube assembly provides information for the final position of the capsule with a high spatial resolution less than 1 mm. The variation of flow rate is also found to indicate the operational status of the pneumatic transfer system. In the case of capsule lost accident, a laboratory scale test has demonstrated that the exact position of the lost capsule can be determined by the sound wave method in which the time delay between an incident sound signal and a reflected one by the capsule is measured so as to provide the position of the capsule with a spatial resolution of 0.2 m. These two capsule position monitoring methods are expected to improve the accuracy, operational stability, and the ability to handle the accident in the ITER NAS.