Neutronic design of Indian LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) TBM to be tested in ITER. Liquid lead lithium along with lithium titanate has been adopted as basic material in Indian TBM for neutron multiplication and tritium breeding. RAFMS is used as the structural material and the first wall is cooled by helium. Li-6 enrichment is taken as 60 and 90% in lithium titanate and lead lithium, respectively. The LLCB TBM design is under progress and two design variants are being considered viz. plate design and tube design. In plate design the lead lithium and lithium titanate zones are arranged alternatively and are parallel to the first wall of TBM. In tube design circular tubes of RAFMS are assumed parallel to first wall and lead lithium flows inside the tubes or outside the tubes and lithium titanate is placed accordingly. For the neutronic design of the LLCB TBM, a detailed 3D neutronic model with “look alike” LLCB TBM in equatorial port in ITER has been constructed. A 3D neutron source has been used for the D-T neutrons emitted by plasma. Neutronic study is carried out using Monte Carlo transport code with FENDL-2.1 library with the following objectives: (1) to examine the profiles of heating and tritium production rates in the LLCB TBM, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanical analysis, and (3) to compare the tritium production capabilities of two variants of the geometries. This paper will present the main findings from this neutronic study.     

Development of a potential based code for MHD analysis of LLCB TBM

A two dimensional solver is developed for MHD flows with low magnetic Reynolds’ number based on the electrostatic potential formulation for the Lorentz forces and current densities which will be used to calculate the MHD pressure drop inside the channels of liquid breeder based Test Blanket Modules (TBMs). The flow geometry is assumed to be rectangular and perpendicular to the flow direction, with flow and electrostatic potential variations along the flow direction neglected. A structured, non-uniform, collocated grid is used in the calculation, where the velocity (u), pressure (p) and electrostatic potential (?) are calculated at the cell centers, whereas the current densities are calculated at the cell faces. Special relaxation techniques are employed in calculating the electrostatic potential for ensuring the divergence-free condition for current density. The code is benchmarked over a square channel for various Hartmann numbers up to 25,000 with and without insulation coatings by (i) comparing the pressure drop with the approximate analytical results found in literature and (ii) comparing the pressure drop with the ones obtained in our previous calculations based on the induction formulation for the electromagnetic part. Finally the code is used to determine the MHD pressure drop in case of LLCB TBM. The code gives similar results as obtained by us in our previous calculations based on the induction formulation. However, the convergence is much faster in case of potential based code.     

Status of Indian LLCB TBM program and R&D activities

The Indian LLCB TBM, currently under development, will be tested from the first phase of ITER operation (H–H phase) in one-half of the ITER port no-2. The present LLCB TBM design has been optimized based on the neutronic as well as thermal hydraulic analysis results. LLCB TBM R&D activities are primarily focused on (i) development of technologies related to various process systems such as Helium, Pb–Li liquid metal and tritium, (ii) development and qualification of blanket materials viz., structural material (IN-RAFMS), tritium breeding materials (Pb–Li, and Li₂TiO₃), (iii) development and qualification of fabrication technologies for TBM system. The present status of LLCB TBM design activities as well as the progress made in major R&D areas is presented in this paper.     

Progress in engineering design of Indian LLCB TBM set for testing in ITER

The Indian Lead–Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is the Indian DEMO relevant blanket module, as a part of the TBM program in ITER. The LLCB TBM will be tested from the first phase of ITER operation in one-half of an ITER port no. 2. LLCB TBM-set consists of LLCB TBM module and shield block, which are attached with the help of attachment systems. This LLCB TBM set is inserted in a water-cooled stainless steel frame called ‘TBM frame’, which also provides the separation between the neighboring TBM-sets (Chinese TBM set) in port no. 2. In LLCB TBM, high-pressure helium gas is used to cool the first wall (FW) structure and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder (CB) pebble bed to cool the TBM internals which are heated due to the volumetric neutron heating during plasma operation. Low-pressure helium is purged inside the CB zones to extract the bred tritium. Thermal-structural analyses have been performed independently on LLCB TBM and shield block for TBM set using ANSYS. This paper will also describe the performance analysis of individual components of LLCB TBM set and their different configurations to optimize their performances.     

3D thermo-fluid MHD simulation of single straight channel flow in LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) blanket for its DEMO fusion reactor. The mock-up blanket (TBM), using this concept, will be tested in ITER for its tritium breeding and high-grade heat extraction efficiency. In this TBM, pressurized helium is used to remove the heat from first wall, top and bottom plates of TBM. The Pb–Li is used to extract heat from the breeder zones. The flow of Pb–Li with average velocity 0.1 m/s inside the channel can be significantly modified due to MHD effects, which arise because of the presence of strong toroidal magnetic field. A numerical approach is established to capture this flow modification at higher Hartmann numbers (≥20,000). As a validation part of the developed code, MHD phenomenon is studied in 2-D square geometry and numerically obtained velocity profile is compared with available Hunt's analytical results. Thermo-fluid MHD analysis using this code, has been carried out for single rectangular duct of LLCB TBM. The heat transfer has been studied by keeping hot breeders at both sides of the flow channel. The results suggest modification in steady state MHD velocity profile as the liquid flows along the flow length. However, the temperature in various zone remains well within the maximum allowable limit.     

Preliminary safety analysis of ex-vessel LOCA for the European HCPB TBM system

Ex-vessel loss of coolant accident caused by a double-ended pipe break of the helium coolant system inside port cell is considered as one of the most critical accident for the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) system. The resulting rapid helium blow-down causes an immediate block of the TBM cooling, which requires a prompt plasma shutdown. Even after the plasma shutdown the temperature can increase over the design limit and the accident sequence can lead up to a break of the TBM box protection after the failure of different protection systems. Thus air ingresses in the vacuum vessel from the damaged TBM system and steam from the surrounding ITER blanket and divertor structures. The evaluation of this sequence is very important for the definition of the correct protection strategy of the system. To consider all these different events a methodology has been developed in KIT combining different codes for a complete analysis of the accident. In particular, this paper shows an application of MELCOR code to model beryllium–steam reaction in a particular accidental sequence for the long term cooling.     

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner.     

用于六氟化铀泄漏事故后果评估的化学物质推荐限值

在六氟化铀泄漏事故后果的评估中,除了关注铀的放射性后果外,还需要关注泄漏物质的化学毒性影响。在进行化学毒性影响评估时,如果采用半致死浓度作为评估标准,则低估了六氟化铀泄漏事故的影响半径。在进行六氟化铀泄漏事故的环境影响评估时,不仅要考虑泄漏物质的致死效应,更应该考虑泄漏物质对事故地点周边公众的健康影响。美国环保署、美国工业卫生协会等机构均提出了相应的化学浓度限值系列。对六氟化铀事故后果评价,推荐采用美国能源部推出的保护行动限值(PACs)。     

Preliminary Shielding Analysis for HCCB TBM Transport

A preliminary shielding analysis on the transport of the Chinese helium cooled ce?ramic breeder test blanket module (HCCB TBM) from France back to China after being irradiated in ITER is presented in this contribution. Emphasis was placed on irradiation safety during trans?port. The dose rate calculated by MCNP/4C for the conceptual package design satisfies the relevant dose limits from IAEA that the dose rate 3 m away from the surface of the package con?taining low specific activity III materials should be less than 10 mSv/h. The change with location and the time evolution of dose rates after shutdown have also been studied. This will be helpful for devising the detailed transport plan of HCCB TBM back to China in the near future.     

中国实验快堆保护系统可靠性评价

核电安全性始终是世界公众关注的焦点。目前概率安全评价已成为核电厂安全评价的标准方法之一,用概率论的方法对核电厂各个系统进行可靠性评价也已成为必不可少的分析手段。中国实验快堆是我国第一座池式钠冷快中子反应堆,其概率安全评价和可靠性评价对其安全运行具有重要意义。本文建立了中国实验快堆保护系统的故障树,并进行了定性和定量分析,计算得出了系统的不可用度,并进行重要度、敏感度和不确定度分析,对典型位置的部件进行共因评价。分析结果表明:由于多个冗余通道的设计,系统最小割集的发生概率分布比较均匀,该系统的设计是平衡的,不存在明显的薄弱环节,能够满足反应堆发生异常事件时的停堆要求。     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.     

The integrated management system as an auxiliary means for providing data for the probabilistic safety assessment

The results of probabilistic safety analyses (PSA) provide plant-specific reliability parameters which reflect more precisely the plant investigated. These parameters are ascertained by a sufficiently extended study of the individual systems and components. An extension of the study period with regard to the total operational life of a plant is to be recommended in most cases. The cost of preparing the required data and processing it to obtaining plant specific reliability parameters for the components can, however, be considerable. The volume and contents of the operational documentation, including maintenance instructions and recurrent tests, are, in all German power stations, more or less the same in view of the fact that these instructions are laid down by law. The accessibility to the data acquired during operation is dependant on the auxiliary means used for the documentation. At the NPP Emsland a computer-aided integrated management system (IBFS) has being used since its start-up. In this paper the methodical procedure for data supply from the IBFS will be explained. In the IBFS all data of operational events are available, since all tasks of plant operation are carried out on the basis of the IBFS. These are optimal prerequisites to obtain from the IBFS the necessary data and information for the PSA-relevant components.     

The hypertext-based reference procedure used in an expert system for life assessment of high-temperature components

The paper presents first results of the development of the hypertext-based reference procedure envisaged to be used in an expert system for assessment and management of the remaining life of the high-temperature high-pressurized components of power and process plants. The development is related to the collaboration among a group of European partners on development of expert system applications in the structural assessment and lifetime prediction and management of high-temperature components. The reference procedure is envisaged as a summary of good practice, recommendations, standard regulatory requirements, human expert experience and case histories. The procedure should appear both as a paperwork and as hypertext based software, running on a personal computer. The hypertext software prototype is presented here in more detail.     

Addressing the challenges posed by advanced reactor passive safety system performance assessment

Passive safety features play an essential role in the development of nuclear technology and within advanced water cooled reactor designs. The assessment of the reliability of such systems in the frame of plant safety and risk studies is still an open issue. This complexity stems from a variety of open points coming out from the efforts conducted so far to address the topic and concern, for instance, the amount of uncertainties affecting the system performance evaluation, including the uncertainties related to the thermal-hydraulic (T-H) codes, as well as the integration within an accident sequence in combination with active systems and human actions. These concerns should be addressed and conveniently worked out, since it is the major goal of the international community (e.g. IAEA) to strive to harmonize the different proposed approaches and to reach a common consensus, in order to add credit to the underlying models and the eventual out coming reliability figures. The main key points that may influence the reliability analysis are presented and discussed and a viable path towards the implementation of the research efforts is delineated, with focus on T-H passive systems.     

高放废物地质处置安全评价准则研究

本文从安全评价的原则和目标出发,研究分析了地质处置安全评价工作开展的步骤和方法,提出了安全评价准则指南建议。     

A critical reappraisal of nuclear power plant safety against accidental aircraft impact

The overall problem of nuclear power plant safety against an accidental aircraft impact is discussed in relation with its structural analysis and design. Associated risks, such as fire, which is a potential source of damage for buildings and other structures, are not considered.The paper is divided in two parts. In part I different approaches used for determining the reaction-time curve are discussed. The influence on the results of target motions is examined next. It is shown that for the evaluation of structural response an aircraft-structure interaction analysis is usually an unnecessary refinement, “mean” reaction-time and impact area-time curves being sufficient to define the excitation. Preliminary results for oblique impact are also given. Since the conditional probability of a normal impact is very small, the consideration of oblique impact may become acceptable in future design criteria.In part II, available solutions for the resulting structural dynamic problem are reviewed. The feasibility of resorting to a static analysis is also discussed. Present practices to evaluate floor response spectra are reviewed next. The short-comings of the “deterministic” approach are pointed out. It is proposed to define the excitation as a mean plus a fluctuating force. The latter is treated as a nonstationary random process and the problem solved by numerical integration in the time domain. Although such solutions get prohibitively expensive when the number of degrees of freedom becomes large, results obtained for simple models may help to clarify which are the important variables of the problem.     

A GoldSim model for the safety assessment of an HLW repository

An assessment program for the evaluation of a high-level waste (HLW) repository has been developed by utilizing GoldSim, by which nuclide transports in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release could be modeled and evaluated. To demonstrate its usability, three illustrative cases including the influence of a groundwater flow pattern through canisters associated with a flowing groundwater through fractures, and the possible disruptive events caused by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical Korean HLW repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, the illustrative cases are very informative especially when seeing the result of the probabilistic calculation with the groundwater flow patterns which is turned out to be possibly important for nuclide release and further transport in and around the repository system.     

Design options for safety depressurization system

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature, a safety depressurization system (SDS) to rapidly depressurize the primary system in case of events beyond the design basis. Two design approaches are considered for the SDS design. The use of bleed valves similar to the ABB-CE System 80+ is design option 1, while in design option 2, the French Sebim valve is considered to provide the combined function of overpressure protection and rapid depressurization. In this paper, thermal hydraulic analysis using a best-estimate version of CEFLASH-4AS/REM is performed for a total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option, various feed and bleed (F and B) procedures are investigated for a TLOFW event. For design option 1, the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with a sufficient margin using design option 1. For design option 2, the operator action times for initiating the F and B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible.