Study on the tritium behaviors in the VHTR system. Part 1. Development of tritium analysis code for VHTR and verification

A tritium permeation analyses code (TPAC) has been developed at Idaho National Laboratory (INL) by using MATLAB SIMULINK package for analysis of tritium behaviors in the VHTR integrated with hydrogen production and process heat application systems. The modeling is based on the mass balance of tritium-containing species and hydrogen (i.e., HT, H₂, HTO, HTSO₄, and TI) coupled with a variety of tritium source, sink, and permeation models. The code includes: (1) tritium sources from ternary fission and neutron reactions with ⁶Li, ⁷Li ¹⁰B, and ³He; (2) tritium purification system; (3) leakage of tritium with coolant; (4) permeation through pipes, vessels, and heat exchangers; (5) electrolyzer for high temperature steam electrolysis (HTSE); and (6) isotope exchange for SI process. Verification of the code has been performed by comparisons with the analytical solutions, the experimental data, and the benchmark code results based on the Peach Bottom reactor design. The results showed that all the governing equations are well implemented into the code and correctly solved. This paper summarizes all the background, the theory, the code structures, and some verification results related to the TPAC code development at INL.     

Numerical prediction of the fission product plate-out for a VHTR application

In order to identify and quantify the key parameters affecting the predictions of fission product transport and plate-out behavior in the coolant circuits of a very high temperature reactor (VHTR) system, systematic and intensive analyses were performed based on numerical predictions as well as the existing experimental data. For the purpose, the computational module for modeling fission product transport phenomena was developed and incorporated into the system analysis code, GAMMA+ for an integrated analysis. This integration can provide more realistic boundary conditions such as velocity, temperature, etc., during off-normal conditions as well as normal operations in a given VHTR system.     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手,对5个循环方案进行比较分析。综合考虑各种工程因素,上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的,其中下位循环在高温气冷堆范围是亚临界参数循环,在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化,且反应堆入口温度在反应堆压力壳材料允许的范围内,具有足够的安全性。     

Safety study of the coupling of a VHTR with a hydrogen production plant

The present paper deals with specific safety issues resulting from the coupling of a nuclear reactor (very high temperature reactor, VHTR) with a hydrogen production plant (HYPP). The first part is devoted to the safety approach consisting in taking into account the safety standards and rules dedicated to the nuclear facility as well as those dedicated to the process industry. This approach enabled two main families of events to be distinguished: the so-called internal events taking place in the coupling circuit (transients, breaks in pipes and in heat exchangers) and the external events able to threat the integrity of the various equipments (in particular the VHTR containment and emergency cooling system) that could result from accidents in the HYPP. By considering a hydrogen production by means of the iodine/sulfur (IS) process, the consequences of the both families of events aforementioned have been assessed in order to provide an order of magnitude of the effects of the incidents and accidents and also in order to propose safety provisions to mitigate these effects when it is necessary. The study of transients induced by a failure of a part of the HYPP has shown the possibility to keep the part of the HYPP unaffected by the transient under operation by means of an adapted regulation set. Moreover, the time to react in case of transfer of corrosive products in the VHTR containment has been assessed as well as the thermohydraulic loading that would experience the coupling pipes in case of very fast uncoupling of the facilities aiming at avoiding an excessive pressurization of the VHTR containment. Regarding the external events, by applying a method used in the process industries, the bounding representative scenarios have been identified on the basis of their consequences but also on the basis of their occurrence frequency. The consequences of the selected bounding scenarios, calculated taking into the source-term, the atmospheric dispersion and the pressure and toxic effects induced respectively by a hydrogen unconfined vapour cloud explosion (UVCE) and a sulfur dioxide release have been assessed. The resulting safety distance of about 100 m for the UVCE is fairly acceptable in terms of performance (head loss and thermal loss) of the coupling system. However, the longer safety distance (about 1.5 km) calculated for a SO₂ release implies to foresee a long distance to settle the control room of the site or to foresee provisions able to stop very fast the SO₂ leak.     

Overview of R&D activities on tritium processing and handling technology in JAEA

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main R&D activities are: the tritium processing technology for the blanket recovery systems; the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.     

Development of a computer code system for the analysis of prism and pebble type VHTR cores

Korea Atomic Energy Research Institute (KAERI) is developing a new computer code system for an analysis of very high temperature gas-cooled reactor (VHTR) cores based on the existing HELIOS/MASTER code system. Several methodologies were developed in order for the original light water reactor (LWR) code system to treat the unique VHTR characteristics easily such as the so-called double-heterogeneity problem, the effects of a spectrum shift and a thermal up-scattering, a strong fuel/reflector interaction, etc. The method of a reactivity-equivalent physical transformation (RPT) and the equivalent cylindrical fuel (ECF) model are proposed to transform the double-heterogeneous fuel problem into a single-heterogeneous one in a cylindrical coordinate for both a prismatic fuel and a pebble-bed fuel. An eight energy group structure with appropriate group boundaries has been constructed in the MASTER diffusion nodal calculation, within which the issues of a spectrum shift and a thermal up-scattering are resolved. The concern about a strong fuel/reflector interaction can be handled easily by applying the equivalence theory to a simple one-dimensional spectral geometry consisting of the fuel and reflector regions. By combining all the methodologies described above, a well-known two-step core analysis procedure has been established, where HELIOS is used for the transport lattice calculation and MASTER for the 3-D diffusion nodal core calculation. The applicability of our code system was tested against several core benchmark problems. The results of these benchmark tests revealed that our code system is very accurate and practical for an analysis of both the prismatic and pebble-bed reactor cores.     

Control strategies for transients of hydrogen production plant in VHTR cogeneration systems

Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design.In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine-Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant.It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.     

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

The tritium concentration in the hydrogen product in Japan's future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO₃ decomposer and a H₂SO₄vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10^?3 Bq/cm³ at STP (38 Bq/g-H₂) to 0.18 Bq/cm³ at STP (2,000 Bq/g-H₂). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10^?5 Bq/cm³ at STP (0.36 Bq/g-H₂) to 5.6 × 10^?3 Bq/cm³ at STP (63 Bq/g-H₂), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10^?2 and 9.6 × 10^?3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.     

A thermal-fluid assessment of a cooled-vessel concept for a VHTR

Flow distribution and thermal analyses of a conceptual design of a cooled vessel for a very high temperature reactor (VHTR), which has a forced vessel cooling with an internal coolant path through a permanent side reflector, have been performed. A computational fluid dynamics (CFD) code was employed to investigate flow distributions at inlet and upper plenums of the proposed cooled-vessel concept. Thermal-fluid analyses of the cooled vessel during a normal operation were carried out by using the CFD code with the boundary conditions provided by the GAMMA system analysis code. The transient analyses during postulated accidents were conducted by the GAMMA code itself. According to the results, the flow deviation at the riser holes due to a change of the inlet flow path to the core inlet is about ±20% which results in about a 3-7% core flow deviation from the average value depending on the upper plenum height. The pressure drops in the inlet and upper plenums are estimated to be from 13 to 25 kPa with a change of the upper plenum height. A cooling flow of more than 4 kg/s is sufficient to maintain the RPV temperature within the required limit during a normal operation. Transient analysis reveals that the reactor vessel is exposed to a temperature above its limit of 371 °C but this duration is shorter than the allowable time for a creep region with a sufficient safety margin. The results suggest that the cooled-vessel concept considered in this paper has the potential to be used for a VHTR but further and more detailed studies are required to realize the proposed concept.     

Influence on moisture and hydrocarbons on conversion rate of tritium in catalytic reactors of fusion-DEMO detritiation system

Thoughtful consideration of abnormal events such as fire is required to design and qualify a detritiation system (DS) of a nuclear fusion facility. Since conversion of tritium to tritiated vapor over catalyst is the key process of the DS, it is indispensable to evaluate the effect of excess moisture and hydrocarbons produced by combustion of cables on tritium conversion rate considering fire events. We conducted demonstration tests on tritium conversion under the following representative conditions: (I) leakage of tritium, (II) leakage of tritium plus moisture, and (III) leakage of tritium plus hydrocarbons. Detritiation behavior in the simulated room was assessed, and the amount of catalyst to fulfill the requirement on tritium conversion rate was evaluated. The dominant parameters for detritiation are the concentration of hydrogen in air and catalyst temperature. The tritium in the simulated room was decreased for condition (I) following ventilation theory. An initial reduction in conversion rate was measured for condition (II). To recover the reduction smoothly, it is suggested to optimize the power of preheater. An increase in catalyst temperature by heat of reaction of hydrocarbon combustion was evaluated for condition (III). The heat balance of catalytic reactor is a point to be carefully investigated to avoid runaway of catalyst temperature.     

ANTARES: The HTR/VHTR project at Framatome ANP

Framatome ANP is developing a very high temperature reactor (VHTR), relying on its previous experience with high temperature reactor concepts, from its participation in the MODUL and the GT-MHR designs. While being a major actor in the nuclear reactor business with proven light water technology, AREVA wishes to be ready to meet the new challenges calling for small grid requirements, high temperature process heat and cogeneration. The Framatome ANP VHTR design for electricity production is based on an indirect cycle coupled to an “off-the-shelf” combined cycle gas turbine. Although direct cycle HTRs are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding power generation system (PGS) developments and keeping the PGS contamination free. Moreover, the nuclear heat source of the indirect cycle could also be used to meet the heat supplies from a standard design for multiple applications.     

Study on oxidation of hydrogen over commercial catalyst for tritium recovery system

For the establishment of the D-T fusion reactor technology, recovery of tritium released into the working area of fusion power plants is quite important. When tritium leaks to working areas, the last barrier is the wall of the building. Due to higher diffusion coefficient of tritium, it diffuses through the wall and would be readily liberated to the environment. Thus, the tritium recovery system is indispensable for the D-T fusion reactor. The objective of the present study is to develop the advanced technology of the tritium recovery system.In the near future, deuterium plasma discharge experiments scheduled be conducted with Large Helical Device (LHD) in National Institute for Fusion Science. A small amount of tritium is produced by D-D reaction in LHD. Tritium in plasma exhaust gases and process gas during discharge needs to be recovered, and thus the design and construction of the tritium recovery system used for that purpose is a matter of considerable urgency.The tritium recovery system usually consists of catalysts and adsorbents, which is the most conventional and reliable method for removing tritium that is accidentally released into the working area of these facilities. However, more recent and advanced type of catalysts on the market cannot be directly applied to the design of tritium recovery system, because of paucity of design data for tritium recovery system. In this study, the authors performed oxidation experiments of hydrogen over a catalyst. The experiments were performed by changing various experimental parameters.     

Simplified optimum sizing and cost analysis for compact heat exchanger in VHTR

In this study, the optimum size of the compact heat exchanger has been developed based on its weight and pumping power for the reference design of 600 MWt very high temperature gas-cooled reactor (VHTR) system. Alloy 617 was selected as a construction material. The optimum size and a number of configurations for the reference design of the VHTR with an intermediate heat exchanger (IHX) were investigated and our initial calculations indicated that it has an unrealistically too large aspect ratio of the length and height due to its small-sized channels, which might cause manifolding problems and a large number of parallel modules with high thermal stress. The flow area and channel diameter were then adjusted to achieve a smaller aspect ratio in this paper. Achievement of this aspect ratio resulted in higher cost, but the cost increase was less than would have occurred by simply reducing the flow area by itself. The appropriate channel diameter is estimated to be less than 5.00 mm for the reference system. The effect of channel waviness enhanced the compactness and heat transfer performance, but unfavorably increased the aspect ratio. Therefore, the waviness should be carefully determined based on performance and economics. In this study, the waviness of the IHX is recommended to be selected between 1.0 and 2.5. Calculations show that reducing the duty dramatically decreases the aspect ratio, indicating that the compact heat exchanger is easy to be optimally designed for low duty, but many modules are required for high duty operation proportional to the operating power. Finally, the effect of working fluids was investigated, and it reveals that using carbon dioxide instead of helium in the secondary side reduces the size and cost by about 20% due to the lower pumping power in spite of its lower heat transfer capability by a factor of 4.     

金属样品残余氚测量方法

为定量评价氚在结构材料和老龄贮氢材料内部的滞留量,用化学蚀刻法测定不锈钢内部氚浓度大小、分布情况及同位素交换后老龄贮氚铀床的氚滞留量。结果表明,贮氚13年的不锈钢样品中氚主要存在于样品内表面由表及里的120μm范围内,样品蚀刻深度110.6μm范围内,不锈钢的平均氚滞留量～9.37×10-4mmol/g,贮氚铀粉平均氚滞留量～4.16×10-5mmol/g。该方法对测量金属中微量氚有较高灵敏度,可检测金属中残余氚的滞留量。     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

Coating Failure of VHTR Reference Fuel by Irradiation

In a series of irradiation examination on VHTR reference fuels, failure fractions of the coated particles have been measured in the acid-leaching method. All of the failure fractions of the coated particles irradiated in the form of compact were under an allowable limit designed for VHTR fuel. Comparing the failure fractions of the particles in the compact and in the loose form, the former fraction was about an order of magnitude as low as the latter, suggesting that the graphite matrix of the compact might have an effect on protection of coating layers. The metallography of the post-irradiation examination revealed irradiation-induced densification of the buffer layer, which caused debonding of this layer from inner PyC layer or occurrence of radial cracks in this layer. It was found that in the limited case the crack extended to SiC layer, then leading to total coating failure.     

Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM

Tritium behavior in the reactor such as production, diffusion and release are accompanied by their adsorption and desorption in graphite materials, which are essential to the safety of high temperature gas cooled reactor (HTGR). In order to study this important issue, hydrogen instead of tritium is experimentally used in this work and justified viable by theory. By performing multiple sets of comparative experiments, the features of hydrogen adsorption and desorption behavior changing by adsorption temperature and time in typical graphites used in HTR-PM (High Temperature Gas Cooled Reactor – Pebble Bed Module), i.e. reflective layer, fuel element and boron carbon bricks, have been observed and analyzed. Furthermore, the adsorption rates of hydrogen in the three materials as above at different conditions are also given. Based on the experimental results, tritium behavior in the HTR-PM was inferred and estimated, which is significant for the further study on the mechanism of tritium transport.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.     

聚变堆氚分析程序TAS1.0的设计与开发

氚是聚变堆的重要燃料之一,对聚变堆氚系统进行分析从而实行有效的氚控制是聚变研究的重要内容之一.在中国系列液态金属锂铅包层聚变堆概念设计研究基础上,利用现代软件工程方法及面向对象技术设计思想,发展了聚变堆氚分析程序TAS1.0,可用于聚变堆氚自持分析、氚燃料管理及氚安全性分析与研究,并可为聚变堆包层及燃料循环系统设计与分析提供技术支持.通过一系列的测试校验,表明了该程序的正确性与有效性.本文主要介绍该程序的系统设计、技术特点与程序测试.     

A quantitative evaluation of reliability of passive systems within probabilistic safety assessment framework for VHTR

This article presents a quantitative evaluation of the reliability of passive systems (RoPS) within the probabilistic safety assessment (PSA) framework for very high temperature reactors (VHTR). VHTRs have unfavorable features in regard to defining a robust failure state. From the viewpoint of PSA, the evaluation of the RoPS as a part of VHTR’s PSA should carefully consider the correct status of a passive system in order to resolve these unfavorable features. This article focuses on the application of multiple states criteria to determine the status of a passive system. Two approaches, i.e., the exceedance probability (EP) model and the stress–strength interference (SSI) model were proposed for the multiple states of the system. A feasibility study has examined the basic features of the proposed approaches by using the reactor cavity cooling system (RCCS) for Korean VHTR. The primary condition for the usefulness of the proposed approaches is that sufficient information should be provided in order to determine the system strengths for the multiple states. With regard to the engineering practice, the EP approach for the multiple states can provide a practical solution concerning the evaluation of the RoPS for VHTR’s PSA.