High Temperature Engineering Test Reactor Research Articles

Investigation on structural integrity of graphite component during high temperature 950 °C continuous operation of HTTR

Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability.To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling.This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950 °C continuous operation, a high temperature continuous operation with reactor outlet temperature of 950 °C for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950 °C continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000 °C. In addition, the programs of surveillance test and ISI using TV camera were introduced.

Validation and application of thermal hydraulic system code for analysis of helically coiled heat exchanger in high-temperature environment

A qualification of the thermal hydraulic system code RELAP5 (reactor excursion leak analysis program) is conducted for the analysis of helically coiled heat exchangers used in a high-temperature environment. A set of closure models based on the past separate effect test are suggested and validated against the measured data in the high temperature engineering test reactor (HTTR). The modified code is then tested for the analysis of a representative nuclear hydrogen production system.The comparison of calculated and measured data with steady-state operation showed that the original heat transfer package in RELAP5 significantly underpredicts the peak heat transfer tube temperature in the intermediate heat exchanger (IHX), particularly when the inlet helium gas temperature exceeds 600 °C. In contrast, the prediction with the suggested model generally agreed well. Lower prediction of peak temperature was observed in outer shell due to the modeling capability in system codes. However, those and the other results from the analysis of the HTTR-IS system revealed that the simplification of the heat structure geometry in IHX with low operating shell temperature is acceptable. In conclusion, this study explores the applicability of the system code to safety analysis of nuclear hydrogen production systems by means of a newly introduced closure model.

Assessment of damage domains of the High-Temperature Engineering Test Reactor (HTTR)

This paper presents an assessment analysis of damage domains of the 30MWth prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA.

Fuel performance under continuous high-temperature operation of the HTTR

The HTTR (high-temperature engineering test reactor) fuel, the first mass-produced HTGR (high-temperature gas-cooled reactor) fuel in Japan, has been fabricated with high quality on the basis of design principle and safety criteria to minimize both as-fabricated failure and on-operating-additional failure of the coated fuel particle. A precise technology for evaluating irradiation performance of the HTTR fuel has been developed by measuring fission gas released from the fuel and constructing the fission gas release model. Through the HTTR operations, including the continuous high-temperature operation at 950 °C for 50 days, the measured fractional release of fission gas from the fuel was less than 1.2×10−8, and a superior irradiation performance of Japanese HTGR fuel has been demonstrated. The HTTR fuel technologies could make a prospect for realization of the VHTR (very high temperature reactor) as a Generation IV nuclear power plant as well as the practical HTGR for the first time. This paper describes the fuel performance under continuous high-temperature operation of the HTTR.

Development plan of high burnup fuel for high temperature gas-cooled reactors in future

Plan and status of research and development (R&D) were described on coated fuel particle (CFP) and fuel compacts for the core of small-sized high-temperature gas-cooled reactor (HTGR) HTR50S at second step of phase I (second core of HTR50S). Specifications of existing CFPs for high burnup (HTR50S2-type-CFPs) were adopted as specifications of CFPs, to reduce R&D. HTR50S2-type-CFPs were fabricated based on technology developed in High Temperature Engineering Test Reactor (HTTR) project. The first irradiation test of HTR50S2-type-CFPs is now being carried out. In addition, R&D for fuel compact with high packing fraction is planned, because volume fraction of UO2 kernel to whole of HTR50S2-type-CFP is rather smaller than that of the HTTR-type-CFP. We would aim to complete the proof of nuclear/thermal design of second core of HTR50S on integrity of fuel, feeding back results of R&Ds and nuclear/thermal design to each other. In addition, we describe outline of R&D plans for core of HTR50S in phase II and practical HTGR in Japan in future, naturally safe HTGR.

Helium leak and chemical impurities control technology in HTTR

Japan Atomic Energy Agency (JAEA) has designed and developed high-temperature gas-cooled reactor (HTGR) hydrogen cogeneration system named gas turbine high-temperature reactor (GTHTR300C) as a commercial HTGR. Helium gas is used as the primary coolant in HTGR. Helium gas is easy to leak, and the primary helium leakage should be controlled tightly from the viewpoint of preventing the release of radioactive materials to the environment. Moreover from the viewpoint of preventing the oxidization of graphite and metallic material, the helium coolant chemistry should be controlled tightly.The primary helium leakage and the helium coolant chemistry during the operation is the major factor in the HTGR for commercialization of HTGR system.This paper shows the design concept and the obtained operational experience on the primary helium leakage control and primary helium impurity control in the high-temperature engineering test reactor (HTTR) of JAEA. Moreover, the future plan to obtain operational experience of these controls for commercialization of HTGR system is shown.

Establishment of control technology of the HTTR and future test plan

The high-temperature engineering test reactor (HTTR) has carried out several demonstration tests to obtain basic data for the design of a future high-temperature gas-cooled reactor (HTGR). In addition, for commercial HTGRs, it also should be demonstrated that the HTGR system can supply stable heat to heat utilization system for the long term. To demonstrate the capability of HTGRs of stable heat supply, the operational experience of the HTTR would be useful. The control characteristics of the HTTR are evaluated by the result of the long-term high-temperature operation performed in 2010. As the result, it is confirmed that control technology of the HTTR is enough to achieve the control response with reasonable damped characteristics and the stable normal operation. In addition, a future possibility of the thermal-load fluctuation and lost test is examined. The data would contribute design of heat utilization system because heat utilization system is required to be a non-safety system. For the future HTGR system design, the requirement, method and result of determination of control constants are discussed. The behavior of control system is described with the characteristics of the HTGR during the past test using HTTR. In addition, this paper proposes a future test plan for demonstration of the reactor stability and control technology against thermal load fluctuation for future HTGR with a heat utilization system design.

Nuclear design study on a small-sized High Temperature Gas-cooled Reactor with high burn-up fuel and axial fuel shuffling

Japan Atomic Energy Agency (JAEA) has started a conceptual design study of a small-sized High Temperature Gas-cooled Reactor (HTGR) with 50MW thermal power (HTR50S) to be deployed in developing countries in the 2020s. The nuclear design of the HTR50S is performed by upgrading that of a High Temperature Engineering Test Reactor (HTTR), which is the Japanese HTGR with 30MW thermal power. In the HTTR design, 12 kinds of fuel enrichment were used to optimize the power distribution. In the previous study of the HTR50S, we succeeded in reducing the number of the fuel enrichment to 3. The present study challenges the nuclear design for effective use of uranium by utilizing high burn-up fuel and axial fuel shuffling, in which a half of the loaded fuel elements is discharged from the core every 2 years and the remains are reloaded. The core burn-up calculations were performed and the nuclear characteristics were confirmed to satisfy the design requirement.

Study of the applicability of CFD calculation for HTTR reactor

Passive heat removal performance of the reactor vessel cavity cooling system (RCCS) is of primary concern for enhanced inherent safety of HTGR. In a loss of forced cooling accident, decay heat must be removed by radiation and natural convection of RCCS. Thus thermal hydraulic analysis of reactor internals and RCCS is powerful means for evaluation of the heat removal performance of RCCS. The thermal hydraulic analyses using CFD computation tools are conducted for normal operation of the high temperature engineering test reactor (HTTR) and are compared to the temperature distribution of measured data. The calculated temperatures on outer faces of the permanent side reflector (PSR) blocks are in fair agreement with measured data. The transient analysis for decay heat removal mode in HTTR is also conducted.

Near term test plan using HTTR (high temperature engineering test reactor)

JAEA has carried out research and development to establish the technical basis of high temperature gas cooled reactors (HTGRs) using HTTR. In order to connect hydrogen production system to HTTR, it is necessary to ensure the stability of plant dynamics when the thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the stable reactor dynamics and to gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted by HTTR to confirm the validity of the calculated temperature coefficient of reactivity. A loss of forced cooling (LOFC) test using HTTR has been carried out to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled.

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30MW and a maximum reactor outlet coolant temperature of 950°C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9MW to 0MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the strength of the neutron source, the larger the minimum reactor power.

Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

To establish the technical basis of HTGR (High Temperature Gas cooled Reactor), the long term high temperature operation using HTTR was carried out in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to establish the technical basis of operation and maintenance by demonstrating the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, it was confirmed that the temperatures and flow rate of primary and secondary coolant were well controlled within sufficiently small deviation against the disturbance by the atmospheric temperature variation in daily. Stability and reliability of the components and facility special to HTGRs was demonstrated through the long term high temperature operation by evaluating the heat transfer performance of high temperature components, the stability performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the long term high temperature operation of HTTR, the technical basis for the operation and maintenance technology of HTGRs was established.

Investigation of chemical characteristics of primary helium gas coolant of HTTR (high temperature engineering test reactor)

The technical basis of helium gas purification control for HTGRs was established by verifying the design of the Primary Helium Purification System (PHPS) of the HTTR by showing that the measured concentrations of impurities of the primary helium coolant were restricted below the criteria of control to protect the graphite oxidation, and that the carburization atmosphere was maintained to keep intact of metallic high temperature components, in the 30-day continuous operation and the 50-day long-term high temperature operation. The analytical model, which was newly established by improving the conventional method that predicted the impurity concentrations conservatively higher than the measured values, predicted the composition of the impurities such as H2, CO, H2O and CO2, which is determined by the temperature dependency of release of impurities during the rated power operation adequately. In contrast, it was revealed that the measured concentration of H2O remarkably decreased while the concentration of CO increased in the primary helium coolant in the long-term high temperature operation.

Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. One of the DECs for the HTGR is the air/water ingress accidents. As the present status of the HTGR, such as the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Agency (JAEA), graphite sleeves hold the fuel compact which consists of the SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and make coolant flow channels. Then, the air/water ingress accident at the HTGR is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO2) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R&D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies.

A preliminary neutronic evaluation of the high temperature nuclear reactor (HTTR) using reprocessed fuel

The High Temperature Engineering Test Reactor (HTTR), a 30MWth, graphite-moderated, helium-cooled reactor constructed by the Japanese government was simulated using reprocessed fuel obtained by UREX+ and spiked with thorium-232 and with depleted uranium. The effective neutron multiplication factor and the nuclear fuel evolution during the burn-up were analyzed. This study was performed using the ORNL SCALE 6.0 code, with CSAS6 and TRITON6 control modules. The results show in a preliminary way that the burn-up of reprocessed fuels in the HTTR core is possible, although the fissile material quantities should be increased while compared with the enrichments of the standard fuel.

SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of high-temperature gas-cooled reactor configurations. This paper describes part of the results of that effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data and the MCNP predictions.

A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

Review of the natC(n, γ) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,γ) and 12C(n,γ) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK•CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.