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1
Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 30% reactor power
Takamatsu, Kuniyoshi
Annals of Nuclear Energy, 106(3), p.71 - 83, 2017/08
The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.
2
Numerical investigation of the random arrangement effect of coated fuel particles on the criticality of HTTR fuel compact using MCNP6
Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji
Annals of Nuclear Energy, 103, p.114 - 121, 2017/05
This study investigated the random arrangement effect of Coated Fuel Particle (CFP) on criticality of the fuel compact of High-Temperature engineering Test Reactor (HTTR). A utility program coupling with MCNP6, namely Realized Random Packing (RRP), was developed to model a random arrangement of the CFPs explicitly for the specified fuel compact of HTTR. The criticality and neutronic calculations for pin cell model were performed by using the Monte Carlo MCNP6 code with an ENDF/B-VII.1 neutron library data. First, the reliability of the RRP model was confirmed by an insignificant variance of the infinite multiplication factor (k$$_{rm inf}$$) among 10 differently random arrangements of the CFPs. Next, the criticality of RRP model was compared with those of Non-truncated Uniform Packing (NUP) model and On-the-fly Random Packing (ORP) model which is a stochastic geometry capability in MCNP6. The results indicated that there was no substantial difference between the NUP and ORP models. However, the RRP model presented a lower k$$_{rm inf}$$ of about 0.32-0.52%$$Delta$$k/k than the NUP model. In additions, the difference of k$$_{rm inf}$$ could be increased as the uranium enrichment decreases. The investigation of the 4-factor formula showed that the difference of k$$_{rm inf}$$ was predominantly given by the resonance escape probability, with the RRP model showing the smallest value.
3
Current R&D status of thermochemical water splitting iodine-sulfur process in Japan Atomic Energy Agency
Kasahara, Seiji; Iwatsuki, Jin; Takegami, Hiroaki; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Onuki, Kaoru; Kubo, Shinji
International Journal of Hydrogen Energy, 42(19), p.13477 - 13485, 2017/05
Current R&D on the thermochemical water splitting iodine-sulfur (IS) process in Japan Atomic Energy Agency is summarized. Reactors were fabricated with industrial materials and verified by test operations: a Bunsen reactor, a H$$_{2}$$SO$$_{2}$$ decomposer, and a HI decomposer. Reactors of industrial materials showed corrosion stability. Demonstration of the test facility verified integrity of process components and stability of hydrogen production. An 8 hours continuous operation of the total IS process was performed in February 2016 with H$$_{2}$$ production rate of 10 L/h.
4
Probabilistic risk assessment method development for high temperature gas-cooled reactors, 4; Use of operational and maintenance experiences with the high temperature engineering test reactor
Shimizu, Atsushi; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 8 Pages, 2017/04
Present paper provides an approach to update PRA parameters using the operational and maintenance experience obtained from the HTTR. Firstly, components subject to investigation are selected with the following criteria; The component has safety function in commercial HTGR, the component is utilized in high temperature-irradiated condition, structure or mechanism of the action for the component is unique, and the component is installed in the HTTR. Secondly, component boundaries are clarified and raw data is collected from maintenance records, monthly surveillance test records, operation and maintenance database, etc. As a preliminary study, selected PRA parameters are updated using Bayesian methods to confirm the effectiveness of the use of the HTTR experience. The results showed that the use of HTTR operational and maintenance data is effective for HTGR reliability database development.
5
Probabilistic risk assessment method development for high temperature gas-cooled reactors, 1; Project overviews
Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 7 Pages, 2017/04
JAEA, in conjunction with Tokyo City University, The University of Tokyo and JGC Corporation, have started development of a PRA method considering the safety and design features of HTGR. The primary objective of the project is to develop a seismic PRA method which enables to provide a reasonably complete identification of accident scenario including a loss of safety function in passive system, structure and components. In addition, we aim to develop a basis for guidance to implement the PRA. This paper provides the overview of the activities including development of a system analysis method for multiple failures, a component failure data using the operation and maintenance experience in the HTTR, seismic fragility evaluation method, and mechanistic source term evaluation method considering failures in core graphite components and reactor building.
6
Benchmark study on realized random packing model for coated fuel particles of HTTR using MCNP6
Ho, H. Q.; Morita, Keisuke*; Honda, Yuki; Fujimoto, Nozomu*; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 8 Pages, 2017/04
The Coated Fuel Particle plays an important role in the excellent safety feature of the High Temperature Gas-cooled Reactor. However, the random distribution of CFPs also makes the simulation of HTGR fuel become more complicated. The Monte Carlo N-particle (MCNP) code is one of the most well-known codes for validation of nuclear systems; unfortunately, it does not provide an appropriate function to model a statistical geometry explicitly. In order to deal with the stochastic media, a utility program for the random model, namely Realized Random Packing (RRP), has been developed particularly for High Temperature engineering Test Reactor (HTTR). This utility program creates a number of random points in an annular geometry. Then, these random points will be used as the center coordinate of CFPs in the MCNP6 input file and therefore the actual random arrangement of CFPs can be simulated explicitly. First, a pin-cell calculation was carried out to validate the RRP by comparing with Statistical Geometry (STG) model of MVP code. After that, the comparison between the RRP model (MCNP) and STG model (MVP) was shown in whole core criticality calculation, not only for the annular core but also for the fully-loaded core. The comparison of numerical results showed that the RRP model and STG model differed insignificantly in the multiplication factor as expected, regardless of the pin-cell or whole core calculations. In addition, the RRP model did not make the calculation time increase a lot in comparison with the conventional regular model (uniform arrangement).
7
Measurement of temperature response of intermediate heat exchanger in heat application system abnormal simulating test using HTTR
Ono, Masato; Fujiwara, Yusuke; Honda, Yuki; Sato, Hiroyuki; Shimazaki, Yosuke; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Iigaki, Kazuhiko; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 5 Pages, 2017/04
Japan Atomic Energy Agency (JAEA) has carried out research and developments towards nuclear heat utilization of High Temperature Gas-cooled Reactor (HTGR) using High Temperature Engineering Test Reactor (HTTR). The nuclear heat utilization systems connected to HTGR will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations continue even if abnormal events occur in the systems. Heat application system abnormal simulating test with HTTR was carried out in non-nuclear heating operation to focus on the thermal effect in order to obtain data of the transient temperature behavior of the metallic components in the Intermediate Heat Exchanger (IHX). The IHX is the key components to connect the HTTR with the heat application system. In the test, the coolant helium gas temperature was heated up to 120 degree C by the compression heat of the gas circulators in the HTTR under the ideal condition to focus on the heat transfer. The tests were conducted by decreasing the helium gas temperature stepwise by increasing the mass flow rate to the air cooler. The temperature responses of the IHX were investigated. For the components such as the heat transfer tubes and heat transfer enhancement plates of IHX, the temperature response was slower in the lower position in comparison with the higher position. The reason is considered that thermal load fluctuation is imposed in the secondary helium gas which flows from the top to the bottom in the heat transfer tubes of the IHX. The test data are useful to verify the numerical model of the safety evaluation code.
8
Probabilistic risk assessment method development for high temperature gas-cooled reactor, 5; Accident progression analysis
Honda, Yuki; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 8 Pages, 2017/04
We have been conducting a source term evaluation method development for high temperature gas-cooled reactors considering structural failures in the major components. This paper present the results of transient analysis for depressurized loss-of-forced cooling accident with ruptures of the cross cut ducts and standpipe, which may be initiated by earthquake. The sequences accounts failures of mitigation systems such as core heat removal by Vessel Cooling System (VCS) and reactor shut down by control rod systems. We will show the effect of mitigation system failure to depressurized loss-of-forced cooling accident in the view point of fuel temperature and natural circulation flow rate which is important for source term evaluation. The major findings obtained in this study showed that multiple failures in mitigation systems for a representative HTGR plant do not aggravate the accident. The result demonstrated that a simplification of event sequence analysis and source term analysis can be achieved with a design fully utilizing the safety characteristics of HTGR.
9
Study of the reduction method of the helium gas leakage from bolted gasket flanged connection for HTGRs
Hamamoto, Shimpei; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 4 Pages, 2017/04
High temperature gas-cooled reactors (HTGRs) use helium as coolant. Because helium tends to leak, welding is often used for joints of pipes and containers. However, the bolt fastening flange is useful for enhancing the maintainability of the industrial plant. If the helium leak characteristic of the bolt fastening flange is clarified and the factor that reduces the leakage of helium can be controlled, it can lead to the reduction of the leak rate of helium. In this study, it was clarified that the temperature difference between the flange surfaces strongly influences the helium leak rate from the flange by experiment using Helium Gas Circulator installed in High Temperature engineering Test Reactor. We also demonstrated that helium leak can be reduced by using this correlation by controlling the flange temperature.
10
Cost performance design for high temperature helium heat transport piping of GTHTR300C and HTTR-GT/H $$_{2}$$ plants
Nomoto, Yasunobu; Horii, Shoichi; Sumita, Junya; Sato, Hiroyuki; Yan, X.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 9 Pages, 2017/04
This paper presents the cost performance design of heat transport piping systems for GTHTR300C plant and HTTR-GT/H $$_{2}$$ plant. Two types of pipe structure are designed and compared in terms of cost performance. Relative to the coaxial double-pipe structure, the insulated single pipe structure is found to have the advantage in overall cost performance considering both the material quantity and the heat loss because it reduces the quantity of steel used for construction. Furthermore it is possible to reduce the heat loss and temperature reduction of hot helium gas by the attachment of the external insulation. The pressure tube made of type-316 stainless steel with high-temperature strength is possible to achieve the same temperature reduction by smaller diameter than that made of 2 1/4Cr-1Mo steel. It contributes to the reduction of the quantity of steel. Specifications of heat transport piping systems for both plants are determined according to these study results.
11
Probabilistic risk assessment method development for high temperature gas-cooled reactors, 2; Development of accident sequence analysis methodology
Matsuda, Kosuke*; Muramatsu, Ken*; Muta, Hitoshi*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (DVD-ROM), 7 Pages, 2017/04
This paper proposes a set of procedures for accident sequence analysis in seismic PRAs of HTGRs that can consider the unique accident progression characteristics of HTGRs. Main features of our proposed procedure are as follows: (1) Systematic analysis techniques including Master Logic Diagrams are used to ensure reasonable completeness in identification of initiating events and classification of accident sequences, (2) Information on factors that govern the accident progression and source terms are effectively reflected to the construction of event trees for delineation of accident sequences, and (3) Frequency quantification of seismically-initiated accident sequence frequencies that involve multiplepipe ruptures are made with the use of the Direct Quantification of Fault Trees by Monte Carlo (DQFM) method by a computer code SECOM-DQFM.
12
Confirmation of feasibility of fabrication technology and characterization of high-packing fraction fuel compact for HTGR
Mizuta, Naoki; Ueta, Shohei; Aihara, Jun; Shibata, Taiju
JAEA-Technology 2017-004, 22 Pages, 2017/03
Confirmation of feasibility of fabrication technology and characterization of the high-packing fraction fuel compact of High Temperature Gas Reactor (HTGR) fuel were carried out. Fuel compacts were fabricated with CFP packing fraction targeted at 33 percent by the same manufacturing condition of HTTR fuel compact. SiC-defective fraction, compressive strength and internal CFP distribution of the compact, important parameters to guarantee its integrity, were evaluated. The high-packing fuel compacts showed as same level of SiC-defective fraction as that of HTTR first loading fuel, 8$$times$$10$$^{-5}$$, and larger compressive strength than the HTTR fuel criteria, 4,900N. The feasibility of fabrication technology and the performance for the high-packing fraction fuel compact was confirmed.
13
Neutronic characteristic of HTTR fuel compact with various packing models of coated fuel particle
Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji
JAEA-Technology 2016-040, 16 Pages, 2017/03
To study the packing effects of the truncated coated fuel particle on the criticality for the High Temperature engineering Test Reactor (HTTR), four alternative models including the truncated uniform model, the non-truncated uniform model, the truncated random model and the non-truncated random model for the arrangement of CFP in fuel compact were used, and the neutronic and criticality calculation were performed by using Monte Carlo MCNP6 code with ENDF/B-VII.1 cross section library. The results showed that the infinite multiplication factors (k$$_{rm inf}$$) in the truncated models were lower than those of the non-truncated models regardless of the uniform or random arrangement, and the four factors in the four-factor-formula showed that the difference of k$$_{rm inf}$$ was mainly attributed to the resonance escape probability. The difference in resonance escape probability is caused by the increase of capture reactions in the resonance region as the influence of spatial-self-shielding-effect. It is because the equivalent kernel diameter of the CFP for the truncated model is smaller than that of the non-truncated model.
14
Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR); FY2015
Department of HTTR
JAEA-Review 2016-036, 95 Pages, 2017/03
The High Temperature Engineering Test Reactor (HTTR) was attained at the full power operation of 30 MW in December 2001 and achieved the 950 degrees of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. This report summarizes activities and results of HTTR operation, maintenance, and several Research and developments, which were carried out in the fiscal year 2015.
15
Development of fuel temperature calculation code "FTCC" for high temperature gas-cooled reactors
Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju
JAEA-Data/Code 2017-002, 74 Pages, 2017/03
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
16
Development of fuel temperature calculation code for HTGRs
Inaba, Yoshitomo; Nishihara, Tetsuo
Annals of Nuclear Energy, 101, p.383 - 389, 2017/03
 Percentile:100
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
17
Development of transportation container for neutron startup source of High Temperature Engineering Test Reactor (HTTR)
Shimazaki, Yosuke; Sawahata, Hiroaki; Yanagida, Yoshinori; Shinohara, Masanori; Kawamoto, Taiki; Takada, Shoji
JAEA-Technology 2016-038, 36 Pages, 2017/02
The High Temperature Engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of $$^{252}$$Cf with 3.7GBq The NSs are exchanged at the interval of approximately 7 years. The NS holders including NSs are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the Control Rod guide blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Following technical issues were extracted from the experiences in the past two exchange works of NSs to develop a safety handling procedure; (1) The reduction and prevention of radiation exposure of workers. (2) The exclusion of falling of NS holder. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container and satisfying the regulation of A type transportation package.
18
Improvement of neutron startup source handling work by developing new transportation container for High-Temperature engineering Test Reactor (HTTR)
Shimazaki, Yosuke; Sawahata, Hiroaki; Shinohara, Masanori; Yanagida, Yoshinori; Kawamoto, Taiki; Takada, Shoji
Journal of Nuclear Science and Technology, 54(2), p.260 - 266, 2017/02
 Percentile:100
The High-Temperature engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of $$^{252}$$Cf with 3.7 GBq and is contained in a small capsule, installed in NS holder and subsequently in a control guide block (CR block). The NSs are exchanged at the interval of approximately 7 years. The NS holders are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container. As the results, the NS handling work using the new transportation container was safely accomplished by developing the new transportation container which can reduce the risks of radiation exposure dose of workers and exclude the falling of NS holder.
19
Shielding calculation by PHITS code during replacement works of startup neutron sources for HTTR operation
Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.
20
Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy
Fukaya, Yuji; Goto, Minoru
Annals of Nuclear Energy, 99, p.19 - 27, 2017/01
 Percentile:100
Sustainable and safe energy supply with seawater fueled HTGR have been investigated to sustain the nuclear energy safely by electricity generation with HTGR, the uranium resources must be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. The uranium dissolved in seawater is in an equilibrium state with the uranium on surface of sea floor, which is approximately a thousand times of the amount, that is 72 million years. It can be recoverable. In other words, the uranium from seawater is almost inexhaustible natural resource. The cost of extracting uranium from seawater with current technology is still expensive compared with that of conventional uranium. However, the economy of nuclear power generation fueled by seawater uranium should be assessed for entire electricity generation cost. In the present study, the economy of electricity generation using uranium from seawater is assessed using a commercial HTGR. Compared with ordinary LWR using conventional uranium, HTGR can realize lower cost of electricity owing to small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR fueled by seawater uranium with the current technology enables the energy sustainability to be maintained for a long term approximately 70 million years with superior inherent safety features and low cost of 7.28 yen/kWh, which is lower than the 8.80 yen/kWh cost of LWR using conventional uranium.