LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.     

Progress in the KIT approach for development of the HCPB TBM stiffening plate feasibility mock up fabrication

Sub-component manufacturing and assembly concepts for the fabrication of the Helium Cooled Pebble Bed Test Blanket Module have been developed since more than one decade in the KIT. In the present design the structure of the HCPB TBM can be sub-divided into three key components: (i) TBM box, (ii) stiffening plates and (iii) the breeder zone. The present fabrication and assembly routines is based on the assumption that each of the aforementioned sub-components can be assembled in parallel and independently before assembling the TBM. Therefore the procedures to fabricate these sub-components can be addressed in independent tasks. This paper shows the results of the KIT/industry collaboration with the final goal to develop a set of preliminary welding procedure specifications (pWPS) for the assembly of the HCPB TBM stiffening plate. Recently a promising set of draft pWPS could be identified in medium scale fabrication experiments. This paper recalls the results of qualification routines according to ISO 15614-11 (RCC-MR Edition 2007, RS 3570) in order to verify the parameters.     

Sensitivity study on in-box LOCA for a Korean HCCR TBM in ITER

Korea has designed a Helium-Cooled Ceramic Reflector (HCCR)-based Test Blanket System (TBS) for International Thermonuclear Experimental Reactor (ITER). Among seven selected reference accidents in Korean TBS, in-box loss of coolant accident (LOCA) is one of them. This is initiated by a double-ended break of the coolant pipe in the Breeding Zone (BZ), pressurizing the BZ box structure, causing pressurization of the Tritium Extraction System (TES) and purging of pipelines. When the accident is detected, the Plant Safety System (PSS) isolates the Helium Cooling System (HCS) and TES, and requests plasma shutdown to Fusion Power Shutdown System (FPSS). To prevent aggravating failure of the system, the safety function is automatically activated when the accident is detected, the device being the isolation valve of HCS and TES. One important observation of this accident is that instant isolation is not a good measure to take. In terms of the possibility of aggravating failure, system isolation is an important safety procedure but isolated TES volume is exposed to high pressure and temperature conditions in the early move of the accident transient. The result of system safety analysis shows that delayed isolation keeps the system safe for a while. In this article, given the preliminary accident analysis results for the current HCCR TBS, case studies were performed regarding the delayed isolation timing effect. For this transient simulation, Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant.In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing.In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.     

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current design of DFLL-TBM and its auxiliary system meets the thermal-hydraulic and safety requirements from ITER.     

Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1

A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions.     

Non-linear failure analysis of HCPB Test Blanket Module

For the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) the reduced activation ferritic martensitic (RAFM) steel EUROFER 97 is selected as a structural material. During operation the TBM will be subjected to complex thermo-mechanical loadings which yield at certain positions of the structure to stresses beyond the design limits of the structural material. Preliminary structural analyses of the TBM have shown critical behavior in several key points of the structure. An improved design has been proposed and in order to identify and assess the problematic positions in the improved version of the TBM a non-linear failure analysis is performed, for which a coupled deformation damage model developed at KIT for RAFM steels and recently implemented in the finite element code ABAQUS is used. The thermal loads in the form of non-homogeneous temperature fields distributions are obtained from a thermal analysis performed using the finite element code ANSYS on the same structure. Importing the temperature fields into the finite element code ABAQUS and applying the remaining loads – coolant internal pressure and structural boundary conditions – non-linear simulations are conducted taking into account the ITER-typical cyclic nature of the loading. The simulation results are evaluated and discussed considering ratcheting and damage at most critical highly loaded areas of the structure.     

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.     

Investigation into the in-box LOCA consequence and structural integrity of the KO HCCR TBM in ITER

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) and its auxiliary system in ITER. In parallel with its design, safety analysis has performed including accident analysis with the selected reference accidents. Among them, the effect of in-box LOCA to the structural integrity of the TBM was investigated. From the transient analysis of the GAMMA-FR on the in-box LOCA, it is found that the pressure of the internal TBM can be increased up to 8 MPa with the same pressure of the operating coolant through the Tritium Extraction System (TES) and He purge lines in the TBM. Structural analysis with ANSYS code for TBM was performed with this condition and it is confirmed that the TBM can endure and it does not affect the ITER machine by the failure.     

Design and performance analysis of ITER ex-vessel magnetic diagnostics

Accurate magnetic diagnostics are essential to perform reliable operation of any tokamak. The ITER magnetic diagnostics include a wide variety of sensors located on the inner and outer surfaces of the vacuum vessel, in the divertor cassettes and in the casing of the toroidal field coils. As the measurement accuracy of the inner set of magnetic sensors might be compromised by various radiation effects and high heat loads, the complementary ex-vessel set is essential to provide backup information. This paper is an overview of the ex-vessel magnetic diagnostic which consists mainly of pick-up coils, steady state sensors, Rogowski coils in the toroidal field coil casing and fibre optic current sensors. The work presented aims at designing these sensors to meet the performance requirements in spite of the constraints due to the tokamak environment. The manufacturing constraints and the positioning requirements for all the ex-vessel magnetic sensors are described. The use and expected accuracy of the entire ex-vessel magnetic diagnostic is assessed in terms of magnetic equilibrium reconstruction and plasma current measurement precision.     

Blowdown force analysis of piping system under LOCA conditions using BLOWDOWN code

The BLOWDOWN code was developed for blowdown force analysis of piping system under LOCA conditions. This is a post-processor of the thermal-hydraulic analysis code RELAP4/MOD6. The results obtained from the RELAP4/MOD6 code are converted into blowdown forces by the BLOWDOWN code.In the paper, the physics and algorithms of the BLOWDOWN code are outlined. Some numerical examples are also presented to show the effectiveness of the code.     

TECAR, a new program for LOCA analysis

A new space- and time-dependent containment pressure and temperature analysis code, TECAR, has been developed. The new code is valid for dry containments, has all the features of the present codes used by industry and in addition calculates the temperature in the containment as a function of height (temperature-space dependence). The code has been verified against the experimental data available.     

HCLL and HCPB coolant purification system: Design of the copper oxide bed

The coolant purification system (CPS) together with the tritium extraction system (TES) and helium cooling system (HCS) are the principal auxiliary circuits of helium-cooled-lithiium–lead (HCLL) and helium-cooled-pebble-bed (HCPB) test blanket modules (TBMs). To extract heat from TBMs, Helium is used as primary coolant. CPS is used to extract tritium from the helium primary circuit as well as to guarantee removal of impurities which could interact with structural material. The reference process proposed for CPS is composed of 3 main successive steps. Step 1 consists in oxidation of Q₂ and CO to Q₂O and CO₂ using a copper oxide bed (Q represents either: H, D or T). Step 2 is dedicated to the removal of water which is adsorbed together with CO₂ on molecular sieve bed. Step 3 will remove residual impurities using a heated getter.Based on the operating conditions of CPS (pressure, flowrate, temperature) and on an estimation of the impurities foreseen, this paper presents a design of the oxidising bed which fulfils all requirements in terms of efficiency and lifespan. The design is obtained using a phenomenological approach taking into account competition between oxidation of CO and Q₂ on the metal oxide. The model was implemented in matlab software. A column of 0.41 m large and 2 m long containing 480 kg of CuO is proposed to assure complete oxidation of Q₂ for 16 months long.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

中国氦冷固态实验包层模块In-box LOCA事故分析研究

中国氦冷固态实验包层模块(CN HCCB TBM)将在ITER 2号窗口进行测试,在测试期间,聚变中子和TBM内部材料发生核反应,产生氚和其他放射性物质。考虑到ITER的运行和工作人员与公众的安全,在进入ITER测试之前需要进行事故安全分析。本文应用MELOCR对HCCB TBM及其氦冷系统(HCS)进行建模,开展了TBM增殖区冷却板流道破口事故(In-box LOCA)安全研究,并对泄压罐体积,破口面积,隔离阀关闭延迟时间等关键参数进行敏感性分析。结果表明:在保守假设流道全破裂的工况下,box压力超过其压力限值4 MPa,而单根流道和5根流道破裂的工况下,box均未超过其压力限值;安装泄压罐和改变隔离阀关闭延迟时间能够有效的控制box压力。     

Structural analysis of ITER TBM Frame and Dummy TBM

One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets. To accomplish these goals, three ITER equatorial ports are dedicated to the test of Test Blanket Modules (TBMs) that are mock-ups of tritium breeding blankets. These TBMs, associated with appropriate shield blocks, will also provide the same thermal and nuclear shielding as the main blanket. The main function of TBM Port Plug (PP) is to accommodate TBMs and provide a standardized interface with the vacuum vessel (VV)/port structure.The feasibility of the design concept of the Frame including two Dummy TBMs has been investigated by proposing design improvements of the reference design through an extensive set of thermal, electromagnetic (EM) and stress analyses. As well, the related static strength was evaluated in accordance with the structural design criteria for ITER in-vessel components (SDC-IC). This paper outlines the engineering aspects of the ITER TBM Frame and Dummy TBM and focuses on the feasibility of the present design by structural assessment.     

Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed （HCPB） Test Blanket Module （TBM） for ITER Using System Code RELAP5

The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.     

船用核动力装置小破口失水事故放射性后果分析

建立了小破口失水事故下热工水力分析与放射性源项计算耦合模型,利用研发的反应堆源项放射性计算软件(Nuclear source radioactive compute,NSRC),分别就不同破口尺寸的堆舱放射性泄漏进行了分析和研究,进一步研究了小破口失水事故,冷端安注和热端安注对堆舱放射性影响。结果表明:破口尺寸大小、安全注射位置及破口隔离时间直接影响堆舱放射性泄漏大小。本工作的分析结果为小型船用堆在小破口设计基准事故下,放射性污染后果分析及事故处置提供了依据。