Language：English  

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Language：English  

Language：English  

Language：English  

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Language：English  

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Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

In a block-type High-Temperature Gas-Cooled Reactor (HTGR), a fixed number of burnable poisons are loaded at the beginning of operation time to control a large amount of excess reactivity. Therefore, it is necessary to evaluate the burnable poison (BP) reactivity worth to achieve the optimum design of the HTGR. In this study, an experiment to measure the burnable poison worth reactivity was conducted at the B-Core of Kyoto University Critical Assembly (KUCA). B-core is solid moderator materials such as polyethylene and graphite combined with fuel plates to form the fuel element. The experiment was performed to measure the reactivity worth of a small cadmium plate (15 × 15 × 0.5 mm) at the B-Core of KUCA. In the experiment, there are 8-unit cells in a fuel element. In this study, the unit cell position of cadmium is called the cadmium unit cell. The experiments were carried out by varying the cadmium plate position inside the cadmium unit cell. This study evaluated the cadmium reactivity worth using the Monte Carlo MVP3. The objective of this study was to evaluate the appropriate results between the measured and the calculation values. The simulation using MVP3 code was conducted by varying the number of batches in the calculation. The results showed that the maximum discrepancy between experimental and calculated results was 24% for 5,000 batches. However, the discrepancy decreased when the number of batches increased to 50,000. The cadmium reactivity worth difference between the experiment and simulation was approximately 18 % depending on the cadmium plate position in the cadmium unit cell.

Scopus

Language：English   Publisher：The Japan Society of Mechanical Engineers  

<p>In a block-type High-Temperature Gas-Cooled Reactor (HTGR), a fixed number of burnable poisons are loaded at the beginning of operation time to control a large amount of excess reactivity. Therefore, it is necessary to evaluate the burnable poison (BP) reactivity worth to achieve the optimum design of the HTGR. In this study, an experiment to measure the burnable poison worth reactivity was conducted at the B-Core of Kyoto University Critical Assembly (KUCA). B-core is solid moderator materials such as polyethylene and graphite combined with fuel plates to form the fuel element.</p><p>The experiment was performed to measure the reactivity worth of a small cadmium plate (15 × 15 × 0.5 mm) at the B-Core of KUCA. In the experiment, there are 8-unit cells in a fuel element. In this study, the unit cell position of cadmium is called the cadmium unit cell. The experiments were carried out by varying the cadmium plate position inside the cadmium unit cell.</p><p>This study evaluated the cadmium reactivity worth using the Monte Carlo MVP3. The objective of this study was to evaluate the appropriate results between the measured and the calculation values. The simulation using MVP3 code was conducted by varying the number of batches in the calculation. The results showed that the maximum discrepancy between experimental and calculated results was 24% for 5,000 batches. However, the discrepancy decreased when the number of batches increased to 50,000. The cadmium reactivity worth difference between the experiment and simulation was approximately 18 % depending on the cadmium plate position in the cadmium unit cell.</p>

DOI： 10.1299/jsmeicone.2023.30.1271

CiNii Research

Language：English   Publisher：The Japan Society of Mechanical Engineers  

<p>The conceptual design of a demonstration hydrogen production facility using heat supply from the high-temperature engineering test reactor (HTTR) is being researched and developed at Japan Atomic Energy Agency (JAEA). This facility produces hydrogen with a thermochemical water-splitting Iodine-Sulphur (IS) process that requires high-temperature heat to extract hydrogen efficiently. The HTTR could supply the heat to the IS demonstration plant through the secondary helium loop, coupling the IS plant to the HTTR.</p><p>The rated operation mode of the HTTR gives helium outlet temperature of 850οC with 660 effective full power days (EFPD). However, in order to achieve hydrogen with high efficiency, the helium outlet temperature should be as high as 950οC. Increasing the outlet temperature increases the reactor core temperature, and as a result the operation time decreases. If the operation time is reduced too much, it is not feasible to use the HTTR as a heat supply for the IS plant. Therefore, the purpose of this study is to estimate the operation time of the HTTR at high operation mode of 950οC to confirm whether it could supply long enough high-temperature heat for the demonstration hydrogen IS plant. The thermal-hydraulic model is also revised using the latest calculation method to improve the accuracy of the temperature distribution in the HTTR.</p><p>As results, the core temperature increase by about 50 to 100οC when the outlet temperature increases from 850 to 950οC. Although the increase of core temperature makes keff decrease by about 0.3 %Δk/k, the HTTR can still operate approximately 660 EFPD. Therefore, it is possible to use the HTTR for long-term high-temperature heat supply to the demonstration hydrogen production IS plant.</p>

DOI： 10.1299/jsmeicone.2023.30.1543

CiNii Research

Language：English  

Language：English   Publisher：Nuclear Engineering and Design  

Estimation of decay gamma distribution in a reactor core is essential for safely conducting various works after reactor shutdown such as periodic maintenance, shuffling fuel, removing spent fuel at the end of cycle, etc. Because of the dependency on the complex operating history of the reactor, attempting to calculate the decay gamma rays distribution in the core remains a challenge. This study shows a method to calculate the shutdown gamma distribution in the HTTR core by coupling a Monte-Carlo transport calculation code MCNP6 and an activation code ORIGEN2 to take advantage of spatial dependence and transport abilities of MCNP6 and the detailed fission products tracking during burnup and cooling of ORIGEN2. As result, the three-dimensional shutdown gamma distribution in the HTTR core for different cooling times and spatial locations could be obtained accurately.

DOI： 10.1016/j.nucengdes.2022.111913

Web of Science

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Language：English  

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Language：English   Publisher：Annals of Nuclear Energy  

Power distribution plays a significant role in preventing the fuel temperature from exceeding the safety limit of 1600 °C in high-temperature gas-cooled reactors. Experiments for measuring the power distribution in the graphite-moderated system were performed at the Very High-Temperature Reactor Critical Assembly (VHTRC) facility. The power distribution was determined from the measured Cu activation rate for both the radial and axial distributions. In this study, the pin-wise power distribution of the VHTRC was evaluated with the Monte Carlo MVP3 code. The difference between the calculated and measured results was less than 1 % for the axial and radial distributions. The significant results were concerned with the area around the fuel and reflector regions in the axial direction, where the average discrepancy between the calculated and measured values was 0.8 %. This result showed improved agreement compared to the diffusion calculation that was conducted in the previous study.

DOI： 10.1016/j.anucene.2022.109314

Web of Science

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Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

During operation of the HTTR, hundreds of technical signals and operating conditions must be observed and evaluated to ensure safe operation of the reactor. The accumulated experiment data of the HTTR is not only important for the HTTR operation, but also for the basic development of the HTGR in general. Artificial intelligence (AI) and particularly machine learning (ML) could give the ability to make predictions as well as allow the extraction of key information about physical process from large datasets. Hence, there is a lot of potential to apply AI and ML to predict the operating and safety parameters of the HTTR. In this study, the control rod position of the HTTR is predicted based on ML without using the conventional neutronic codes. The ML with a linear regression algorithm finds a functional relationship between the input dataset and a reference dataset, constructing a function that predicts control rod position from the other operation conditions. As result, the ML gives a good prediction of the HTTR control rod position with less than 5% difference compared to that in the experiment. With increasingly complicated experiments that create a large amount of data, ML is also expected to improve the design and safety analysis of the HTTR in the future.

DOI： 10.1115/ICONE29-90818

Scopus

Publisher：International Conference on Nuclear Engineering, Proceedings, ICONE  

The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-To-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35 % than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pincell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4 %.

DOI： 10.1115/ICONE29-90755

Scopus

Language：English  

DOI： 10.1016/j.anucene.2021.108270

Language：English  

DOI： 10.1016/j.nucengdes.2021.111161

Language：English  

DOI： 10.1115/1.4044529

Language：English  

Language：English  

The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors (HTGRs) was investigated. A high-temperature engineering test reactor (HTTR), which is located in Japan at Oarai-machi Research and Development Center, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the
¹²⁵
I production could be maximized and the
¹²⁶
I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 1.8 × 10
⁵
GBq/y of
¹²⁵
I production.

DOI： 10.1016/j.apradiso.2018.07.024

Language：English  

The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor (HTGR) constructed in Japan, firstly. The operating data of the HTTR with burn-up is very important for developments of HTGRs. Many test data have been collected in the HTTR. Many tests are carried out in low power operation. On the other hand, the full power operation is not enough. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate the calculation code. Additionally, it is difficult to measure core temperature in HTTR. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle and the temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for confirm the characteristics of burn-up and validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.

DOI： 10.1115/1.4033812

Language：English  

The Coated Fuel Particle plays an important role in the excellent safety feature of the High Temperature Gascooled 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).

Language：English  

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation
than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

DOI： 10.1115/1.4033813

Language：English  

The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.

DOI： 10.1080/00223131.2015.1054910

Language：English  

The High Temperature Engineering Test Reactor (HTTR) [1] is a block type fuel High Temperature Gas-cooled Reactor. There are 32 control rods (16 pairs) in the HTTR. The 6 pairs of control rods are inserted into a core region and the others are inserted in a reflector region surrounding the core. The core temperature of the HTTR is too high to insert all control rods simultaneously at reactor scram near full power operation for keeping integrity of control rods metallic sleeve. Therefore, a two-step control rods insertion method for reactor scram is adopted. The reactivity inserted at the two-step control rod insertion method was measured at HTTR criticality tests. The calculated reactivity at the firststep showed larger underestimation than that of the second-step. On the other hand, calculated results of excess reactivity at the HTTR criticality tests showed good agree with tests. It is considered that a cell model for reflector region control rod is not suitable. Therefore, this paper focuses on a new cell model for control rods in a reflector region. In a previous control rod cell model, control rod is surrounded by fuel blocks only. The surrounding condition of the new cell model corresponds to the configuration around the reflector region control rod. The calculated reactivity at the first-step using the new cell model shows better results than previous calculation. It is considered that the new cell model brings appropriate neutron flux distribution around control rods in reflector region.

Language：English  

Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the high-temperature engineering test reactor (HTTR). Additional measurements of the subcritical configuration of the fully loaded core, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely detailed models of the HTTR as much of the design information is still proprietary. The uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. However, use of the benchmark critical configurations of the HTTR for nuclear data adjustment is not recommended as the impact of these biases has not been addressed with rigorous detail. The impact of any simplification biases, if any, is not expected to significantly impact evaluation of the other reactor physics measurement calculations. Dominant uncertainties in the experimental k_eff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that compose the HTTR. Monte Carlo calculations of k_ff are between ∼0.9&#37; and ∼2.7&#37; greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulations of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

DOI： 10.13182/NSE14-14

Language：English  

The Japan Atomic Energy Agency has been establishing a database of operation and maintenance experience for the High Temperature Engineering Test Reactor. The objective of this database is to share information from operation and maintenance experience and make use of the knowledge gained in the design, construction, and operation of future High Temperature Gas-cooled Reactors (HTGRs). Between 1997 and 2012, more than 1000 events have been registered in this database system.
This paper describes trends in operation and maintenance events recorded in this database, including experience gained from the Great East Japan Earthquake. The paper also identifies the following significant items that are expected to be useful in the design of future HTGRs: (1) performance degradation of helium gas compressors, (2) malfunction of the reserved shutdown system in the reactivity control system, (3) problems with emergency gas turbine generators, and (4) consequences of the Great East Japan Earthquake.

DOI： 10.1080/00223131.2014.946568

Language：English  

Benchmark models were developed to evaluate six cold-critical and two warm-critical, zeropower
measurements of the high-temperature engineering test reactor (HTTR). Additional measurements
of the subcritical configuration of the fully loaded core, core excess reactivity, shutdown margins, six
isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as
acceptable benchmark experiments. Insufficient information is publicly available to develop finely
detailed models of the HTTR as much of the design information is still proprietary. The uncertainties in
the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias
uncertainties incurred through the simplification process used to develop the benchmark models.
However, use of the benchmark critical configurations of the HTTR for nuclear data adjustment is not
recommended as the impact of these biases has not been addressed with rigorous detail. The impact of
any simplification biases, if any, is not expected to significantly impact evaluation of the other reactor
physics measurement calculations. Dominant uncertainties in the experimental keff for all core
configurations come from uncertainties in the impurity content of the various graphite blocks that
compose the HTTR. Monte Carlo calculations of keff are between *0.9&#37; and *2.7&#37; greater than the
benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulations of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

Language：English  

In the primary cooling system of the High Temperature Engineering Test Reactor (HTTR) with an outlet coolant temperature of 950°C, high-temperature components and piping such as an intermediate heat exchanger and coaxial double piping reach very high temperature, and large and complex thermal displacements arise in them. In order not only to absorb the thermal displacements but also to withstand earthquakes, the HTTR has adopted a new three-dimensional floating support system. In the limited space of the containment vessel, the support system can support the components and piping's own weights and follow the thermal displacements and have seismic capacity. On the other hand, the adoption of the support system was unprecedented in nuclear plants. Thus, the effectiveness of the support system was demonstrated through the HTTR operation. In this paper, by using the HTTR operation data, the thermal displacement behavior of the high-temperature components and piping is investigated, and the behavior and characteristics are simulated numerically. In addition, the aftermath of the Great East Japan Earthquake on the HTTR is confirmed. As a result, the effectiveness of the three-dimensional floating support system adopted by the HTTR is verified.

DOI： 10.1080/00223131.2014.967734

Language：English  

A high-temperature gas-cooled reactor (HTGR) is a graphite-moderated and helium gas-cooled reactor. It is particularly attractive due to its capability of producing high-temperature helium gas, and its passive and inherent safety features. To enable nuclear energy application to a wide range of heat process industries, Japan Atomic Energy Agency (JAEA) has continued extensive effort for the development of the HTGR using the high-temperature engineering test reactor (HTTR), which is the first HTGR in Japan with a thermal power of 30 MW, and operates it at the site of the JAEA's Oarai Research and Development Center. The HTTR has successfully completed a full-power high-temperature (950°C) continuous operation for 50 days from January to March in 2010. Through this operation, the potential of a stable high-temperature heat supply to heat application systems, such as a hydrogen production system, was demonstrated. This paper presents the operation results including reactor characteristics.

DOI： 10.3327/taesj.J11.020

Language：Japanese  

Language：English  

In block-type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained shallow to keep fuel temperature below 1495 ◦C through a burnup period, and hence excess reactivity should be reduced through a different method. Loading burnable poisons (BPs) into the core is considered as a method to resolve this problem as in case of light water reactors (LWRs).
Effectiveness of BPs on reactivity control in LWRs has been validated by experimental data, however, this has not been done yet for HTGRs, because there was not enough burnup characteristics data for HTGRs required for the validation. The High Temperature Engineering Test Reactor (HTTR) is a block-type HTGRs and it adopts rod-type BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate effectiveness of rod-type BPs on reactivity control in the HTTR, we investigated on the HTTR results whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then effectiveness of rod-type BPs on reactivity control in the HTTR was validated.

Language：English  

Comparison of burnup characteristics of the High Temperature Engineering Test Reactor (HTTR) were conducted between experimental data and analytical results, and then applicability of a three dimensional whole core burnup calculation method to burned block-type HTGRs was studied. The calculations were performed by the SRAC/COREBN based on diffusion theory. We focused on changes in control rods (CRs) position with burnup as one of the HTTR burnup characteristics, and analyzed that in rated power with 30MW and zero power, respectively. In the HTTR experiments, different tendencies were observed in the changes in CRs position with burnup between rated and zero power. CRs position in rated power decreased with burnup, namely, insertion depth of CRs into a core increased with burnup. Meanwhile that in zero power was almost constant through burnup. In the analysis, similar results to the above were obtained. Thus, applicability of the three dimensional whole core burnup calculation method to burned block-type HTGRs was confirmed by experimental data.

Language：English  

High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Agency (JAEA) achieved the reactor outlet coolant temperature of 950°C for the first time in the world at Apr. 19, 2004. To ensure the thermal integrity of fuel in high temperature test operation, it is necessary that fuel temperature is designed appropriately by fuel temperature designing method, and that estimated maximum fuel temperature is lower than the thermal limit temperature. In this report, by constructing newly a realistic core-shape representing model, the current fuel temperature estimation model is improved. Moreover fuel temperature in high-temperature test operation is estimated with the newly-constructed model, and it is confirmed that estimated maximum fuel temperature in high temperature test operation is lower than the thermal limit temperature.

DOI： 10.3327/taesj2002.5.57

Language：English  

Annular cores were formed in start-up core physics tests of the High Temperature Engineering Test Reactor (HTTR) to obtain experimental data for verification of design codes. The first criticality, control rod (CR) positions at critical conditions, neutron flux distribution, excess reactivity, etc., were measured as representative data. These data were evaluated with the MVP Monte Carlo code, which can consider directly the heterogeneity of coated fuel particles (CFPs) distributed randomly in fuel compacts. It was made clear that the heterogeneity effect of CFPs on k_eff's for annular cores is smaller than that for fully loaded cores. The measured and the calculated k _eff's agreed with each other with differences <1&#37;Δk. The calculated neutron flux distributions agreed with the measured results. A revised method was applied for evaluation of excess reactivity to exclude the negative shadowing effect of CRs. The revised and calculated excess reactivity agreed with differences <1&#37;Δk/k.

DOI： 10.13182/NSE03-79

Language：English  

This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations.

DOI： 10.1016/j.nucengdes.2004.08.015

Language：English  

The high temperature engineering test reactor is the first block-type HTGR designed for a 950 °C outlet gas temperature which uses low-enriched uranium fuel with burnable poison rod. For validation of the nuclear design code system for the HTTR, a critical assembly of VHTRC had been constructed. The calculation uncertainties of effective multiplication factor, neutron flux distribution, burnable poison reactivity worth, and control rod worth, temperature coefficients were evaluated. Calculation accuracy of a Monte Carlo code is also evaluated.

DOI： 10.1016/j.nucengdes.2004.08.005

Language：English  

The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950 °C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600 °C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.

DOI： 10.1016/j.nucengdes.2004.07.008

Language：English  

From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.

DOI： 10.1016/j.nucengdes.2004.08.014

Language：English  

The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88&#37; of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal-hydraulic design gives conservative results.

DOI： 10.1016/j.nucengdes.2004.07.009

Language：English  

The high temperature engineering test reactor (HTTR) is a graphite-moderated and helium-gas-cooled reactor, which is the first high temperature gas-cooled reactor in Japan. The HTTR achieved its first full power of 30 MW at rated operation on December 7 in 2001. In the rise-to-power test of the HTTR, simulation test of anticipated operational occurrence with scram was carried out by manual shutdown of off-site electric power from 30 MW operation. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, mass flow rates of helium and water decreased to the scram points. Sixteen pairs of control rods were inserted at two-steps into the core by gravity within the design criterion of 12 s. In 51 s after the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. In 40 min after the startup of the auxiliary cooling system, one of two auxiliary helium circulators stopped for reducing thermal stresses of core graphite components such as fuel blocks. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. Blackout sequences of the HTTR dynamic components were in accordance with the design. As a result of the loss of off-site electric power simulation test, it was confirmed that the HTTR shuts down safely after the scram.

DOI： 10.1080/18811248.2002.9715285

Language：English  

The HTTP has achieved the first criticality on 11/10/98. The fuels were loaded into the core from outer region to inner region to obtain characteristic data of annular cores. The annular core is expected to be a core type of future HTGR. Fuel loading schedule was planned based on preliminary calculations by Monte Carlo method. These calculations predicted the first criticality at 16 ± 1 columns. However, the reactor achieved the first criticality at 19 columns. The first criticality was re-predicted by comparing measured and calculated 1/M curves. The calculated 1/M curves were obtained for different critical mass adjusting some parameters such as the amount of impurities, etc. The method is called "1/M sandwich method". This method well predicted the number of fuel columns at the first criticality. The combination of this method with Monte Carlo calculation was a rational method for predicting the first criticality. It was confirmed that Monte Carlo calculation could be used for the evaluation of HTTR with < 1&#37; of error for fully loaded core.

DOI： 10.3327/jaesj.42.458

Language：English  

High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium-cooled reactor which has 30 MW of thermal power and 950°C of outlet coolant-gas temperature. The fuel loading of the HTTR was started on July 1, 1998, from the core periphery. The first criticality was attained in annular type core of 19 columns on Nov. 10, 1998. The startup core physics tests consisted mainly of tests for licensing and tests for establishing the technology bases necessary for HTGRs. It was confirmed in the former tests that all fuel blocks are loaded in certain positions and the excess reactivity is less than the limit. Experimental data for the annular core are obtained in the latter tests. Also, it was confirmed that the inverse kinetics method and delayed integral counting method are useful for the measurement of scram reactivity even if it takes about 10s for rod insertions. Furthermore, control rod worth curve, axial neutron flux distribution, etc. were measured to grasp the core performance. All tests planned in the startup core physics tests had been successfully performed and were completed on Jan. 21, 1999. It was confirmed from the tests that the HTTR was capable to step up to the power ascension tests.

DOI： 10.3327/jaesj.42.30

Language：English  

DOI： 10.3327/jaesj.41.35

Language：English  

A concept for a new reactor system is developed where weapons-grade plutonium can be made worthless for weapons use. It is a pebble bed-type high-temperature gas-cooled reactor that uses plutonium burner ball and thorium breeder ball fuels. The residual amount of ²³⁹Pu in spent plutonium balls becomes <1&#37; of the initial loading. The power coefficient is made negative by reducing the parasitic neutron absorption reaction rate of ¹³⁵Xe.

DOI： 10.13182/NSE97-A24460

Language：English  

The high-temperature engineering test reactor has been designed whose outlet gas temperature is 950°C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600°C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. Control rod destruction of the optimized power distribution was avoided by limiting the depth of insertion. The insertion depth of the control rods is limited by reducing the excess reactivity of the whole core by the burnable poisons to the minimum value necessary for operations.

DOI： 10.13182/NSE96-A24156

Language：English  

DOI： 10.3327/jaesj.38.304

Language：English  

Japan Atomic Energy Research Institute has started the development of the high temperature engineering test reactor (HTTR), a graphite-moderated, helium gas-cooled reactor with 30 MW thermal power and maximum outlet coolant temperature of 950 °C. This paper describes the core thermal and hydraulic (T/H) design procedure, including the validation of the computer code system, design criteria pertaining to the fuel design limit and the evaluated core T/H charateristics. The core T/H design of the HTTR has been carried out considering the specific characteristics of the core structure and the fuel based on R&D results. The coolant flow rate and temperature distribution are evaluated by the flow network analysis code flownet. The fuel temperature distribution is evaluated by the fuel temperature analysis code temdim with multi-cylindrical model using hot spot factors. Fuel design limit for anticipated operational occurrences and fuel temperature limit for normal operation are specified at 1600°C and 1495°C, respectively based on experimental results. Several design considerations are also adopted to realize a high reactor outlet coolant temperature of 950°C. As a result of core T/H design, the effective core flow rate and maximum fuel temperature during the high temperature test operation are 88&#37; and 1492°C, respectively.

DOI： 10.1016/0029-5493(94)90084-1

Language：English  

Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30MW in thermal output and outlet coolant temperature of 850 °C for rated operation and 950 °C for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests.
JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R&D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation.

Language：English  

In the high temperature gas-cooled reactors (HTGRs), the radial and axial heterogeneity resulted from a combination of fuel rods, burnable poison rods, block end graphite and so on causes local power peakings which increase the fuel temperature locally. An method was developed for calculating the local power and the fuel temperature distributions. This method deals with all heterogeneity effects of a whole core in the radial and axial directions with a design code system including a vectorized 3-dimensional diffusion code. The uncertainty of the method had been evaluated through the analyses of the power distribution obtained by critical experiments with the Very High Temperature Reactor Critical Assembly (VHTRC). The difference was less than 3&#37; between the calculated and measured power distributions. From the results, it was confirmed that this method could predict the local power distribution of the HTGR with high accuracy. This method was applied to the evaluation of the fuel temperature of the HTTR. It was shown that the maximum fuel temperature would be lower than the design limit of 1,495°C for the normal operation and that of 1,600°C for the anticipated operational transients.

DOI： 10.1080/18811248.1994.9735115

Language：English  

High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950°C in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly.

DOI： 10.1080/18811248.1993.9734606

Language：English  

This report presents the optimization result with respect to the spatial power distribution of the High Temperature engineering Test Reactor (HTTR) core to achieve a high outlet coolant gas temperature of 950° C. At first, the power distribution optimization procedure was developed to achieve a high outlet coolant gas temperature while maintaining the fuel temperature as low as possible. Secondarily, the optimization procedure thus developed was applied for the power distribution design of the HTTR core. The maximum nominal fuel temperature was reduced about 300°C through the optimization and was 1,321°C. By the power distribution optimization, the maximum fuel temperature was maintained less than the fuel temperature design limit of 1,600°C, even accounting for the temperature increase at the hot spot and the anticipated operational occurrences.

DOI： 10.1080/18811248.1992.9731553

Language：English  

Very High Temperature Gas Cooled Reactor (VHTR) is one of future reactors for multi-purpose use. The reactor core has large amount of graphite moderator with much heat capacity. So the response of the core outlet gas temperature has very long time constant for disturbances. In this report, the control system design has been studied for the start up, the shut down and the load follow operations using PID controllers.

Event date： 2023.11

Language：Japanese  

Venue：茨城県東海村   Country：Japan  

Event date： 2023.11

Language：Japanese  

Venue：茨城県東海村   Country：Japan  

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Event date： 2023.10

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：Sendai, Japan   Country：Japan  

Event date： 2023.10

Language：Japanese  

Venue：Sendai, Japan   Country：Japan  

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Event date： 2023.9

Language：English   Presentation type：Oral presentation (general)  

Venue：愛知県名古屋市千種区 名古屋大学東山キャンパス   Country：Japan  

Event date： 2023.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：東京都   Country：Japan  

Event date： 2022.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2022.12

Language：Japanese  

Venue：オンライン   Country：Japan  

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Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2022.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福岡県福岡市   Country：Japan  

Event date： 2022.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福岡県福岡市   Country：Japan  

Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

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Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

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Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

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Event date： 2022.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

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Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：茨城県日立市   Country：Japan  

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：茨城県日立市   Country：Japan  

Event date： 2022.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：茨城県日立市   Country：Japan  

Event date： 2022.9

Language：Japanese  

Venue：茨城県日立市   Country：Japan  

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Event date： 2022.9

Language：Japanese  

Venue：茨城県日立市   Country：Japan  

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Event date： 2022.9

Language：Japanese  

Venue：茨城県日立市   Country：Japan  

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Event date： 2022.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2021.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2021.12

Language：Japanese  

Venue：オンライン   Country：Japan  

Event date： 2021.12

Language：Japanese  

Venue：オンライン   Country：Japan  

Event date： 2021.12

Language：Japanese  

Venue：オンライン   Country：Japan  

Event date： 2021.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2021.11

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2021.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2021.3

Language：English   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2021.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2021.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2020.12

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福岡県福岡市   Country：Japan  

Event date： 2020.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2020.9

Language：English   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2020.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：オンライン   Country：Japan  

Event date： 2020.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福島県福島市   Country：Japan  

Event date： 2020.3

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2020.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福島県福島市   Country：Japan  

Event date： 2019.12

Language：Japanese  

Country：Japan  

Event date： 2019.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2019.9

Language：English   Presentation type：Oral presentation (general)  

Venue：富山県富山市   Country：Japan  

Event date： 2019.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：茨城大学   Country：Japan  

Event date： 2019.3

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2018.9

Language：Japanese  

Venue：岡山大学   Country：Japan  

Event date： 2018.7

Language：Japanese   Presentation type：Oral presentation (general)  

Country：Japan  

Event date： 2017.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2017.12

Language：Japanese  

Country：Japan  

Event date： 2017.12

Language：Japanese  

Venue：福岡県福岡市   Country：Japan  

Event date： 2017.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：札幌   Country：Japan  

高温ガス炉燃料の核種生成・崩壊挙動への影響評価の一つとして、高温ガス炉用のORIGENライブラリを作成し、温度及び濃縮度の影響評価を行った。その結果、核種の生成量などに対して濃縮度の影響は小さく、減速材温度の影響が大きいことを確認した。

Event date： 2017.9

Language：English   Presentation type：Oral presentation (general)  

Country：Japan  

This study investigated the impact of truncated coated fuel particles (CFPs) on neutronic characteristic of the fuel in a statistical geometry (STG) model. Calculation results showed that the truncated CFPs make the multiplication factor decrease by about 0.1 – 1.0 &#37;∆k/k depended on packing fraction, uranium enrichment, and particle size.

Event date： 2017.5

Language：English   Presentation type：Oral presentation (general)  

Venue：Makuhari Messe Chiba   Country：Japan  

The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor (HTGR) constructed in Japan, firstly. The operating data of the HTTR with burn-up is very important for developments of HTGRs. Many test data have been collected in the HTTR. Many tests are carried out in low power operation. On the other hand, the full power operation is not enough. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate the calculation code. Additionally, it is difficult to measure core temperature in HTTR. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle and the temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for confirm the characteristics of burn-up and validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.

Event date： 2017.5

Language：English   Presentation type：Oral presentation (general)  

Venue：Makuhari Messe Chiba   Country：Japan  

The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor. There are 32 control rods (16 pairs) in the HTTR. The 6 pairs of control rods are inserted into a core region and the others are inserted in a reflector region surrounding the core. The core temperature of the HTTR is too high to insert all control rods simultaneously at reactor scram near full power operation for keeping integrity of control rods metallic sleeve. Therefore, a two-step control rods insertion method for reactor scram is adopted. The reactivity inserted at the two-step control rod insertion method was measured at HTTR criticality tests. The calculated reactivity at the first- step showed larger underestimation than that of the second- step. On the other hand, calculated results of excess reactivity at the HTTR criticality tests showed good agree with tests. It is considered that a cell model for reflector region control rod is not suitable. Therefore, this paper focuses on a new cell model for control rods in a reflector region. In a previous control rod cell model, control rod is surrounded by fuel blocks only. The surrounding condition of the new cell model corresponds to the configuration around the reflector region control rod. The calculated reactivity at the first-step using the new cell model shows better results than previous calculation. It is considered that the new cell model brings appropriate neutron flux distribution around control rods in reflector region.

Event date： 2017.4

Language：English   Presentation type：Oral presentation (general)  

Venue：Kyoto   Country：Japan  

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).

Event date： 2017.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：神奈川県平塚市 東海大学 湘南キャンパス   Country：Japan  

高温ガス炉燃料の核種生成・崩壊挙動へ影響評価一つとして、中性子ペクトルを確認するために、高温ガス炉用ORIGENライブリを作成し解析行った。その結果 、Pu等の生成量に対する中性子スペクトルの影響がことが明らかになった 。

Event date： 2016.9

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：福岡県久留米市   Country：Japan  

HTTRでは、燃焼による過剰反応度の変化を、BPを用いることにより小さく維持している。この燃焼による過剰反応度の変化を低温臨界状態での制御棒位置変化で評価した。解析により、臨界制御棒位置に対するBP周りのメッシュ分割の効果を明らかにした。

Event date： 2016.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：宮城県 仙台市 東北大学   Country：Japan  

高温ガス炉燃料の核種生成・消滅挙動の評価を目指して、高温ガス炉用ORIGENライブラリを作成し解析を行った。軽水炉用ライブラリの結果と比較すると、PuやFP量に差が見られた。

Event date： 2015.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：茨城県水戸市   Country：Japan  

Event date： 2015.3

Language：Japanese   Presentation type：Oral presentation (general)  

Venue：茨城県水戸市   Country：Japan  

Language：Japanese  

DOI： 10.11484/jaea-technology-2020-003

Language：Japanese  

DOI： 10.11484/jaea-technology-2019-012

Language：Japanese  

Language：English  

これまで高温工学試験研究炉の崩壊熱は、軽水炉のデータを基にしたShure の式やORIGEN計算で評価してきたが、厳密には軽水炉の中性子スペクトルと異なることから最適な評価方法を検討する必要がある。このため、黒鉛減速材量を変えた炉心の中性子スペクトルを用い、ORIGEN2コードで崩壊熱及び生成核種を計算して軽水炉の崩壊熱曲線と比較した。

Language：English  

The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2009.

Language：Japanese  

HTTRの機器配管の支持方式には、3次元の浮動支持方式を採用しているため、その熱変位挙動・特性を把握することは、3次元浮動支持方式の熱変位特性を調べるうえでも重要である。HTTRの定格運転では、1次冷却設備の熱変位特性を知るために、高温配管の熱変位測定試験を実施した。本報は、その試験結果及び解析評価結果を示したものである。熱変位挙動は温度変化に対して1次的に変化することを確認した。解析評価では、支持構造物の抵抗力が機器の熱変位挙動に影響を与えることが明らかとなり、また、その抵抗力を最適化することで実測熱変位挙動を再現できることを確認した。

Language：English  

The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2008.

Language：Japanese  

Language：Japanese  

HTTRは1998年の初臨界達成後、定格運転及び高温試験運転の出力上昇試験を経て、現在、供用運転を行っている。今回、HTTRでは長期にわたって熱利用系に安定な熱供給ができることを実証するために、定格・並列運転で30日連続運転を行った。本報はその運転で得られたHTTRの長期連続運転の特性をまとめたものである。

Language：Japanese  

Language：English  

The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2007.

Language：Japanese  

「HTTR運転データベース」は将来型高温ガス炉の実用化開発や高温工学試験研究炉(HTTR)の運転管理に資することを目的にHTTRの運転データを蓄積・整理したデータベースである。対象データは基本的にHTTRの運転より得られた過剰反応度,炉心又はプラント内の各部温度,冷却材中不純物濃度等の測定値を整理・評価したデータである。データベースは重要度の高い運転データを長期的及び系統的に管理する目的で、将来型高温ガス炉の実用化開発やHTTRの運転管理に関する目的別のデータベースから成る構造とした。本報ではHTTR運転データベースのうちHTTR共通データベース,HTTR核特性データベース,ヘリウム純度管理データベース,炉内構造物健全性データベース及びその他データベースの具体例を示す。

Language：Japanese  

我が国初の高温ガス炉であるHTTRでは炉心支持黒鉛構造物の健全性及び特性等を確認するために、供用期間中検査(ISI)としてTVカメラを用いた炉心支持黒鉛構造物の目視検査及びサーベイランス試験片を用いた物性値の測定を行うこととしている。このうちサーベイランス試験では、HTTRの炉内支持黒鉛構造物について高速中性子照射,酸化等による物性値,強度等の経年変化を調べることにしており、ここで得られるデータは、HTTRの炉心支持黒鉛構造物の健全性確認に用いられるほか、第4世代原子炉システムの1つとして国際的に検討を行っている超高温ガス炉(VHTR)の黒鉛構造物の設計等に活用することのできる貴重なものとなる。本報は今後実施することになるHTTRのサーベイランス試験片の炉内への装荷位置及び使用前の状態における物性データをまとめたものである。

Language：Japanese  

HTTRの核特性評価手法は、設計に用いた解析手法をもとに臨界試験で得られた結果に基づいて開発が行われてきた。開発では、まず反応度調整材(BP)反応度価値の評価手法の改良,制御棒挿入孔からの中性子ストリーミング効果の考慮,燃料セルモデルの改良が行われた。これらの改良により、過剰反応度の測定値に対して1&#37;&#36;&#36;Delta&#36;&#36;k/k以下の誤差で一致するモデルを作成することができた。この改良をもとに、出力運転解析のためのモデルの拡張を行った。ここでは、おもにBPの燃焼変化の挙動について濃縮度依存性,温度効果等を検討した。燃焼挙動についての測定データとの比較では、ほぼ一致する結果を得ることができた。また、モンテカルロコードによる出力分布の比較では、両者はほぼ同じ分布を示すことを明らかにした。また、HTTRの燃料体からの&#36;&#36;gamma&#36;&#36;線測定結果との比較では、改良したモデルは測定結果と良い一致を示すことを明らかにした。さらに、設計時に用いられた核設計手法と比較することにより、今回のモデルの改良が、実効増倍率だけでなく出力分布の改善にも効果があることを明らかした。

Language：Japanese  

「HTTR運転データベース」は将来高温ガス炉の実用化開発や高温工学試験研究炉(HTTR)の運転管理に資することを目的にHTTRの運転データを蓄積・整理したデータベースである。対象データは基本的にHTTRの運転より得られた過剰反応度,炉心又はプラント内の各部温度,冷却材中不純物濃度等の測定値を整理・評価したデータである。データベースは重要度の高い運転データを長期的及び系統的に管理する目的で、将来高温ガス炉の実用化開発やHTTRの運転管理に関する目的別のデータベースから成る構造とした。本報ではHTTR運転データベースの全体概要及び作成方針について報告する。また、本データベースのうちHTTR共通データベース,HTTR核特性データベース及びヘリウム純度管理データベースの一部を例として示す。

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HTTRの炉容器冷却設備(Vessel cooling system: VCS)は工学的安全施設の一つであり、配管破断や減圧事故時のような強制循環による炉心の冷却ができない場合に、炉心の熱を圧力容器の周りに設置した水冷管パネルで輻射及び自然対流によって除去する設備である。また、通常運転時には1次遮蔽体のコンクリート温度を制限値以下に保つ機能を有している。これまでの運転で、VCSによる除熱量の変化はないものの、VCSを流れる冷却水の温度レベルが徐々に上がりはじめた。さらに冷却水の温度上昇に伴い、遮蔽体であるコンクリート温度の上昇も懸念された。このまま運転を続けると、コンクリート温度が制限値に近づくことも予想されたため、VCS冷却水の温度レベルを低下させる対策を行うこととした。VCS冷却水の温度を管理している機器の1つにVCS冷却器がある。このVCS冷却器の伝熱性能を評価しその低下の程度を明らかにした。その結果、伝熱性能の低下が認められたため、伝熱管の洗浄を行い伝熱性能の回復を図った。これにより、VCS除熱量を変化させることなく、VCS冷却器の伝熱性能を大きく改善し、VCS冷却水の温度を低下させることができた。

Language：Japanese  

「DELIGHT」は、炉心計算等に必要な群定数を作成する高温ガス冷却炉用格子燃焼特性解析コードである。円環状または球状の高温ガス炉燃料を対象に衝突確率法による格子計算を行う。高温ガス炉燃料特有の被覆燃料粒子による燃料格子の二重非均質性を考慮した燃焼計算が可能なことが特徴として挙げられる。今回、従来のDELIGHTコードをより燃焼度の高い炉心の解析に対応させることを目的に、核データライブラリのJENDL-3.3への更新,燃焼チェーンを詳細化する等の改良を行った。また、可燃性毒物(BP)格子計算モデルにおいて、BP棒周辺の物質領域を多領域化し、BP格子計算の計算精度の向上を図った。本報は、改良DELIGHTコード(DELIGHT-8)の改良点と使用方法について説明するものである。

Language：Japanese  

高温ガス炉の固有の安全性を定量的に実証するため、高温工学試験研究炉(High Temperature Engineering Test Reactor: HTTR)において、反応度投入及び炉心除熱量減少を試験として実機の原子炉で生じさせる安全性実証試験を実施している。安全性実証試験の1つである制御棒引抜き試験について、1点炉近似モデルにより試験時の動特性解析を実施した。実測値と解析値の比較から、1点炉近似モデルが試験の結果を再現できることを確認した。また、添加反応度,温度係数,物性値等の各パラメータについて、制御棒引抜き事象に対する原子炉動特性への感度を明らかにした。

Language：Japanese  

HTTR炉心の燃料最高温度の評価においては、炉心出力密度分布の予測精度向上が重要であり、炉心管理コードとしても用いられる拡散燃焼計算モデルの改良を図る必要がある。拡散計算によるHTTR炉心の出力密度分布解析について、可燃性反応度調整材(BP)を燃料体内に均質に分布させたモデル(BP混合モデル)とBP領域を分離したモデル(BP分離モデル)の解析結果を、グロス&#36;&#36;gamma&#36;&#36;線による出力密度分布測定結果及び連続エネルギーモンテカルロ計算コードMVPの計算値と定量的に比較した。その結果、BP混合モデルでは、炉心の軸方向出力密度分布に対する予測精度が不十分であること、BP分離モデルを用いることにより、予測精度が大幅に改善されることがわかった。

Language：Japanese  

HTTRは、原子炉出口冷却材温度950&#36;&#36;^{circ}C&#36;&#36;の達成を目指した高温試験運転による出力上昇試験を平成15年度に計画している。高温試験運転の実施にあたっては、被覆粒子燃料を使用し、ヘリウムガス冷却を行う我が国初の高温ガス炉であることを念頭に、これまで実施してきた出力上昇試験(定格運転30MW及び原子炉出口冷却材温度850&#36;&#36;^{circ}C&#36;&#36;までの試験)での知見をもとに計画する。高温試験運転においては、温度の上昇に従ってより厳しくなる、原子炉の核熱設計,放射線遮へい設計及びプラント設計が適切であることを確認しながら実施する。本報では、HTTRの安全性確保に重要な燃料,制御棒及び中間熱交換器について、定格運転モードでの運転データに基づき、高温試験運転時の安全性の再確認を行った結果を示すとともに、これまでに摘出された課題とその対策を示した。加えて、高温試験運転における試験項目摘出の考え方を示し、実施する試験項目を具体化した。その結果、原子炉施設の安全を確保しつつ、原子炉熱出力30MW,原子炉出口冷却材温度950&#36;&#36;^{circ}C&#36;&#36;の達成の見通しを得た。

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HTTRは1次冷却材の温度が最高約950&#36;&#36;^{circ}C&#36;&#36;に達するため、制御棒の金属材料には特殊合金としてAlloy800Hが使用されている。この制御棒の使用に関しては、Alloy800Hの強度データにより、制限温度を900&#36;&#36;^{circ}C&#36;&#36;以下と定め、これを超えるような環境で使用された場合には必要に応じて制御棒を交換することになっている。制御棒温度が900&#36;&#36;^{circ}C&#36;&#36;を超える可能性のある事象として高温試験運転からの商用電源喪失に伴う原子炉スクラムが挙げられる。本書では、出力上昇試験で得られた商用電源喪失試験時の実測データを用いて制御棒温度解析を実施した結果を示す。解析の結果、繰り返し使用を考慮した事象の中で制御棒温度が最も高くなる商用電源喪失が発生したとしても、制御棒温度は制限値を上回ることはなく、健全性が確保されることを確認した。

Language：English  

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28,1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is a first Reactor Cavity Cooling System applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it is confirmed that the VCS heat removal at 30 MW power operation is higher than 0.3MW. This paper shows outline of the VCS and test results on the VCS performance.

Language：Japanese  

高温工学試験研究炉の出力上昇試験; 試験経過及び結果の概要

中川 繁昭; 藤本 望; 島川 聡司; 野尻 直喜; 竹田 武司; 七種 明雄; 植田 祥平; 小嶋 崇夫; 高田 英治*; 齋藤 賢司; et al.

JAERI-Tech 2002-069, 87 Pages, 2002/08

JAERI-Tech-2002-069.pdf:10.12MB

高温工学試験研究炉(High Temperature engineering Test Reactor : HTTR)の出力上昇試験は、30MW運転時に原子炉出口冷却材温度が850&#36;&#36;^{circ}C&#36;&#36;となる「定格運転」モードでの試験として、平成12年4月23日から原子炉出力10MWまでの出力上昇試験(1)を行い、その後、原子炉出力20MWまでの出力上昇試験(2),30MW運転時に原子炉出口冷却材温度が950&#36;&#36;^{circ}C&#36;&#36;となる「高温試験運転」モードにおいて原子炉出力20MWまでの出力上昇試験(3)を行った。定格出力30MW運転達成のための試験として平成13年10月23日から出力上昇試験(4)を開始し、平成13年12月7日に定格出力30MWの到達及び原子炉出口冷却材温度850&#36;&#36;^{circ}C&#36;&#36;の達成を確認した。出力上昇試験(4)については、平成14年3月6日まで実施し、定格出力30MWからの商用電源喪失試験をもって全ての試験検査を終了して使用前検査合格証を取得した。「定格運転」モードにおける原子炉出力30MWまでの試験結果から、原子炉、冷却系統施設等の性能を確認することができ、原子炉を安定に運転できることを確認した。また、試験で明らかとなった課題を適切に処置することで、原子炉出力30MW,原子炉出口冷却材温度950&#36;&#36;^{circ}C&#36;&#36;の達成の見通しを得た。

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現在HTTRでは出力上昇試験を進めており、これまで50&#37;出力を達成している。HTTRの出口温度は950&#36;&#36;^{circ}C&#36;&#36;と高いため、出力上昇の過程で炉心内の温度変化が大きい。このような炉心の解析精度の向上を目的として各出力での臨界制御棒位置及び温度係数について測定を行い、解析との比較を行った。解析は、熱流動解析コードと拡散計算のくり返しにより求めた炉内温度分布を用いて、モンテカルロ計算と拡散計算により行った。その結果、臨界制御帽位置はモンテカルロ計算により50mm以下の誤差で一致し、100&#37;出力では2900mm程度になると予想された。温度係数は拡散計算の結果とよく一致した。今後、出力100&#37;までの測定を行い、解析結果と比較することにより解析精度の向上を目指す。

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HTTRの炉心内の情報を得ることを目的として、炉心から燃料体を取り出し再装荷する過程での燃料体からの&#36;&#36;gamma&#36;&#36;線の測定を行った。測定は、燃料体が通過する床上ドアバルブに設置したGM管及びCZT半導体検出器と、スタンドパイプ室に設置したエリアモニタで行い、炉内のウラン濃縮度配分の対称性を考慮して4カラムの燃料体計20体について行った。測定の結果GM管及びCZT検出器による測定では、各カラムでの軸方向の相対分布は解析とほぼ一致したが、炉心上部では解析値が高く、炉心下部では低くなった。エリアモニタによる測定でも軸方向の分布を測定することができた。さらにカラム間の比較も行った。今後は測定結果について詳細な解析・評価を行い、炉内出力密度分布等の評価精度の向上に役立てる予定である。

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これまで、HTTRの炉心解析モデルについて、VHTRCの実験結果を用いて検証が行われてきた。またモンテカルロコードとの比較に基づき、ゼブラ型反応度調整材の形状及び位置の効果、中性子ストリーミング効果を考慮できるようモデルの改良が進められてきた。さらにこの改良モデルを用いて臨界試験の予備解析が行われてきた。しかしながら臨界試験の結果から、予備解析に用いたモデルでも過剰反応度を過大に評価することが明らかとなった。検討の結果、燃料セルの外径が過大で実際より減速材の黒鉛が多いため柔らかい中性子スペクトルとなり、&#36;&#36;^{235}&#36;&#36;Uの核分裂断面積を大きく評価していることが原因であると考えられた。そこで、燃料セルの外径をこれまでより小さい、燃料棒のピッチによる値とすることにより、臨界試験結果とよく一致する結果を得ることができた。

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VHTRC炉心に複数本の反応度調整材(BP)棒を装荷した実験結果について、BP反応度価値の解析精度を評価した。その結果、HTTRの核設計に用いている、ブロック内を均質としたモデルではBP反応度を20&#37;程度過小評価することが明らかとなった。この結果は、BP反応度を系統的に過小評価しているため、過剰反応度を高めに評価するという観点からは保守的であり安全上問題ない。しかしながら、高温ガス炉の設計の合理化、将来炉の設計、HTTRの運転管理等のためには評価精度の向上が必要である。そこで、BP位置をモデル化し炉心内のインポータンス分布をより詳細に考慮すれば精度が向上すると考え、燃料体内でのBP棒の位置を考慮できるよう燃料体のメッシュ分割数を増やしたモデルを作成した。このモデルとともに、炉心計算のBP棒領域に対応する範囲で均質化することにより作成した実効断面積を用いることにより、10&#37;以下の誤差でBP反応度を評価できることが明らかとなった。

Language：English  

The first Research Coordination Meeting for the Coordinated Research Program on the HTTR benchmark problems were held in August 1998. The results and calculation models of JAERI and Forshcungszentrum Jiilich GmbH (FZJ) by diffusion calculation were compared. Both results showed a good agreement at fully-loaded core but the results of JAERI showed about 1 &#37;Ak higher value during fuel loading state. To investigate the cause of the difference, effects of energy group number, neutron streaming from control rod insertion holes and cell models of burnable poison (BP) were studied. As the results, we found that the difference caused by energy group number and neutron streaming were small. The effect of BP cell model was evaluated by sensitivity analysis of dimension of BP cell. Improvements for each calculation model were proposed.

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高温工学試験研究炉(HTTR)の臨界試験の事前評価として、連続エネルギー法に基づくモンテカルロ計算コードMVPにより核特性解析を行った。拡散理論による炉心計算では直接モデル化が困難であった、燃料コンパクト、燃料棒、燃料棒挿入孔、反応度調整材等の燃料体内の非均質構造、制御棒及び制御棒挿入孔、後備停止系ほう素ペレット落下孔、炉心構成要素間の間隙等を詳細にモデル化した。解析により、初回臨界は16カラム前後燃料を装荷した状態で到達する見込みであること、その際第1,2,3リング制御棒を全引き抜きし中心制御棒だけを操作することで臨界調節が可能であることを確認した。また、臨界時の制御棒位置、過剰反応度、炉停止余裕等を求めた。これらの解析結果を臨界試験の計画策定に用いた。

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本報は、HTTR核特性解析コードシステムの炉心解析モデルの改良と、このモデルを用いて行った臨界試験の予備解析結果について報告するものである。解析モデルは、BPの軸方向装荷パターンがゼブラ状であること並びに燃料体内での径方向位置をモデル化できるよう及び制御棒挿入孔等からのストリーミングを考慮できるよう改良した。予備解析では、燃料装荷に伴う実効増倍率の変化、中性子検出器の応答確認、逆増倍係数、制御棒反応度価値、炉停止余裕、動特性パラメータ、中性子束分布及び出力換算係数に関する解析を行った。本報に示した結果は、既に試験計画及び使用前検査に用いている。今後は、この結果と臨界試験結果を比較し、モデル及び試験結果の妥当性の確認を行う計画である。

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高温工学試験研究炉(HTTR)の過剰反応度を燃料追加法によって測定する場合について、制御棒の干渉効果が過剰反応度に与える影響を評価した。制御棒が全引き抜き状態の実効増倍率から求める過剰反応度に比べて、制御棒操作を考慮することによって、-10&#37;~+50&#37;程度の測定値が変化することがわかった。また、干渉効果の影響を小さくするためには、被測定制御棒、補償制御棒とも複数の制御棒を用いればよく、(1)被測定制御棒として第3リング制御棒を除く13対を用い、そのうちの1対の反応度測定の際にその他の12対を補償制御棒として用いる組合わせ、(2)第1リング制御棒6対を(1)と同様に用いる組み合わせ、が過剰反応度測定に適していることが明らかになった。

Language：English  

Language：Japanese  

本報は、高温ガス炉臨界実験装置(VHTRC)の臨界時の実効増倍率、反応度調整材反応度、ボイド反応度の実験結果を汎用中性子・光子輸送計算モンテカルロコード(MVP)により評価し、MVPを高温ガス炉の核特性評価使用する場合の解析精度の評価を行ったものである。解析の結果、臨界時の実効増倍率、反応度調整材反応度、ボイド反応度の解析誤差は最大で、それぞれ0.8&#37;&#36;&#36;Delta&#36;&#36;k,7&#37;,25&#37;以下であった。臨界時の実効増倍率を十分な精度で予測できることを明らかにした。よって、HTTRの炉心特性評価にMVPを適用することが可能であることがわかった。

Language：Japanese  

高温工学試験研究炉(HTTR)の照射性能の向上及び炉心性能を実規模高温ガス炉のもと同等にすることを目的とし、核熱設計の観点から炉心の高性能化の検討を行った。本検討より、ダルマ落とし燃料交換方式を採用することにより、90GWd/tという高い燃料燃焼度を達成できる見通しを得た。また、一体型燃料コンパクトを用いることにより燃料の徐熱性能を向上でき、炉心平均出力密度を7.1W/cm&#36;&#36;^{3}&#36;&#36;まで大きくしても、原子炉出口冷却材温度950&#36;&#36;^{circ}&#36;&#36;Cを達成できる見通しを得た。高速中性子束及び熱中性子束の最大値は、HTTRの約2.2倍及び約1.5倍まで増大できる見込みを得た。炉心有効流量の低下を防止するため、炭素複合材を被覆管に用いた制御棒の使用及び上部遮蔽体ブロックの側面にはめあい構造を設けカラム間の漏れ流れを低減することが、有効であることが分かった。

Language：Japanese  

ペブルベッド型高温ガス炉に兵器級Puを装荷した場合、温度係数が正になるという問題があった。そこで、温度係数を負にする方法を、兵器級Puを劣化U等の親物質と混合せずに、全炉心に兵器級Puの燃料のみを装荷する炉心を対象として検討した。検討の結果、熱領域に共鳴捕獲反応を持つ核種(Er)を炉心に添加すれば、温度係数が負になることを確認した。また、この場合には燃料球の燃焼度が低下するが、Erを添加せずとも、Puの装荷量を増大して熱中性子束のピークを小さくすれば、燃焼度の低下なしに温度係数を負にできることを明らかにした。燃焼度の増大とともに熱中性子束のピークが再び大きくなる可能性があるが、ペブルベッド型炉では燃料球の最大燃焼度の1/2程度で原子炉が平衡状態になるため、初期の装荷量が多い限り熱中性子束のピークが再び大きくなることはない。

Language：Japanese  

HTTRは原子炉出口温度が950&#36;&#36;^{circ}&#36;&#36;Cと高く、このためスクラムに伴う原子炉停止においては、制御棒の高温における繰り返し使用による寿命の低下を避けるため、まず反射体領域の制御棒を挿入して原子炉を未臨界にし、ついで炉心温度が所定の温度(原子炉出口温度が750&#36;&#36;^{circ}&#36;&#36;C)に下がるのを待って、あるいは所定の時間(2400秒)をおいて燃料領域へ制御棒を挿入して常温で未臨界を維持する制御棒2段階挿入方式を採用している。本報告では、2段階挿入方式を用いたスクラム時において燃料領域の制御棒が挿入されるまでの間、原子炉を未臨界に維持できることの確認を行った。この結果、もっとも厳しい条件となる原子炉出口温度950&#36;&#36;^{circ}&#36;&#36;Cからのスクラム時でも0.7&#37;&#36;&#36;Delta&#36;&#36;k/k(制御棒1対のスタックを考慮した場合)の炉停止余裕を確保できることがわかった。

Language：Japanese  

高温工学試験研究炉(HTTR)の炉心支持板は、炉心及び炉心支持黒鉛構造物の鉛直方向の荷重を直接支持する機能を有し、その上部にある炉床部断熱層により、炉心内の高温冷却材(約950&#36;&#36;^{circ}&#36;&#36;C)からの熱伝導を低減するとともに、その下面を低温冷却材(約400&#36;&#36;^{circ}&#36;&#36;C)で冷却して、制限温度を超えない構造としている。炉心支持板下面の冷却材流路には、1次ヘリウム配管、補助ヘリウム配管及び多数の支持板支持柱等の構造物がある。これらの構造物は、冷却材を偏流させる可能性があり、その結果として炉心支持板にホットスポットが生じる可能性がある。炉心支持板下面の冷却材の流動を明らかにするために、3次元熱流体解析コードSTREAMを用いて解析を行なった。更に、その解析結果から得られた流速分布より、炉心支持板の温度分布を解析した結果、ホットスポットが発生するような偏流が生じないことを確認した。

Language：Japanese  

高温工学試験研究炉では、原子炉スクラム時に制御棒被覆管の高温における繰り返し使用による寿命の低下を避けるため、まず反射体領域の制御棒を挿入し、その後炉心温度が所定の温度以下となった時点で燃料領域の制御棒を挿入し、低温まで未臨界を維持する2段階挿入法を採用している。炉心領域制御棒の挿入は、タイマーによる設定時間または原子炉出口冷却材温度の設定値に達した時点で行う。本報は、種々のスクラム条件下での制御棒被覆管温度解析の手法、条件及び結果についてまとめたものである。

Language：Japanese  

本報は、高温工学試験研究炉(HTTR)の動特性解析コードASURAについて、その概要と検証解析についてまとめたものである。ASURAコードはHTTRプラントシステムの制御動特性解析を目的としているため、各種制御系を含めたプラントシステム全体をモデル化している。検証条件は、各種パラメーターサーベイ及びBLOOST-J2コード、THYDE-HTGRコードとのクロスチェック解析を行った。その結果、ASURAコードの特徴及び妥当性が確認された。また、Fort St.Vrain炉での実験データによる3検証も行い、妥当性の確認を行った。

Language：Japanese  

本報は、高温工学試験研究炉(HTTR)について、原子炉圧力容器入口から、炉心入口部までの冷却材の熱流動解析についてまとめたものである。HTTRでは、原子炉圧力容器に流入した冷却材は、炉心と原子炉圧力容器の間を上方へ流れて上部プレナムへ至り、上部プレナム内で反転して下降流となり炉心へ流入する。本報では、冷却材が炉心と原子炉圧力容器の間を上昇する際の冷却材温度上昇及び温度上昇誤差の評価、上部プレナム内における冷却材の3次元熱流動解析による冷却材温度混合の評価についてまとめたものである。また、炉心入口温度の燃料最高温度評価に及ぼす影響についても検討を加えた。

Language：Japanese  

本報は、高温工学試験研究炉(HTTR)の炉心熱流力設計の主要項目である燃料最高温度評価の用いる工学的安全係数についてまとめたものである。工学的安全係数は、系統的性質を有するシステマティック因子と統計的性質を有するランダム因子とに分け、それぞれHTTRの特徴を踏まえて、考慮すべき主項目と値を決定した。

Language：Japanese  

本報は、現在日本原子力研究所が設計を進めている高温工学試験研究炉(熱出力30MW、原子炉出口冷却材温度950&#36;&#36;^{circ}&#36;&#36;C)の炉心熱流力設計の概要と設計結果についてまとめたものである。炉心熱流力設計では、通常運転時及び運転時の異常な過渡変化時において、それぞれ燃料最高温度の制限値を定め、これを超えないように設計することとしている。このため、炉心及び燃料の構造上の特徴を踏まえ、十分な炉心冷却材流量を確保するとともに燃料最高温度を極力低くするように設計する。本設計において得られた燃料最高温度は、基準炉心の燃焼330日における原子炉出口冷却材温度950&#36;&#36;^{circ}&#36;&#36;C運転時の1495&#36;&#36;^{circ}&#36;&#36;Cである。

Language：Japanese  

Language：Japanese  

本報は、高温工学試験研究炉(HTTR)の炉心熱流力設計の基礎となる炉心内冷却材流量配分計画と評価の結果を、解析用データとともにまとめたものである。HTTRの炉心は、黒鉛ブロックを積重ねた積層構造となっており燃料体ブロック及び制御棒案内ブロック内の計画された流路以外に冷却材の流れる流路が構成される、そのため、炉心の有効な冷却の確保のために、このような計画外の流量で極力低減し、冷却材出口温度950&#36;&#36;^{circ}&#36;&#36;C達成のため適切な流量配分を定めている。

Language：Japanese  

本報は、高温工学試験研究炉(HTTR)の設計において、炉心の伝熱流動、特に燃料ブロック内の冷却材流路間の流量配分、燃料ブロック応力解析用熱的境界条件の決定並びに燃料ブロック内の冷却材流路閉塞事故時の温度評価に使用する熱流動・熱伝導連成コードFLOWNET/TRUMPの検証結果について報告するものである。

Language：Japanese  

本報は、高温工学試験研究炉の炉心熱流力設計において、燃料温度解析に使用する計算コードTEMDIMの検証結果についてまとめたものである。検証解析は、HENDEL燃料体スタック実証試験部1チャンネル試験装置による伝熱流動試験結果を用いて行い、燃料温度評価手法の妥当性、保守性が確認された。

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Number of peer-reviewed articles in foreign language journals：2

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Number of peer-reviewed articles in foreign language journals：2

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Number of peer-reviewed articles in foreign language journals：5

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Number of peer-reviewed articles in foreign language journals：8

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Number of peer-reviewed articles in foreign language journals：8

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Type：Competition, symposium, etc. 

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Number of participants：700

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平成31年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

平成30年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

平成29年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

Lecture and discussion at a seminar of Soomkmyung Women's University in present and future status of energy of world and Japan, utilization of nuclear energy.

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