Shigeaki Nakagawa

Development of an Evaluation Model for the Thermal Annealing Effect on Thermal Conductivity of IG-110 Graphite for High-Temperature Gas-Cooled Reactors

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.

Development of a Simulation Model and Safety Evaluation for a Depressurization Accident Without Reactor Scram in an Advanced HTGR

Abstract It is important to use analyses to prove outstanding inherent reactor safety during a severe accident in order to convince the public and licensing authority of the safety advantage of the high-temperature gas-cooled reactor (HTGR). In this study, the simulation of a depressurization accident without reactor scram (DAWS) was performed for a future HTGR with 450-MW thermal output, introducing the annular core of pin-in-block-type fuel, which was originally designed in Japan. The DAWS has the possibility of becoming one of the severe accidents postulated in the HTGR. To perform an accurate simulation, a new analytical model for reactor dynamics and indirect decay heat removal from the surface of the reactor pressure vessel (RPV) during the DAWS was developed. The features of the new simulation model are as follows: 1. A single-channel model is coupled with a two-dimensional reactor thermal model in the new simulation model. The reactor kinetics with a single-channel model during the DAWS is simulated taking into account heat removal from the reactor calculated in the R-Z reactor thermal model, including the RPV and indirect vessel cooling system. No conventional calculation codes with a single channel have a heat removal model from an RPV or were able to simulate precisely the transient during DAWS. 2. A xenon buildup and decay model for the reactivity calculation is made in addition to one point-kinetics approximation to simulate a recriticality and a power oscillation following the initiation of the DAWS. 3. A transient simulation can be performed for two kinds of core models of pin-in-block- and multihole-type fuels. The accurate evaluation of xenon density and core temperature is of prime importance in the simulation of the DAWS. From the simulation result with a proper safety margin, it was confirmed that the safety performance of passive decay heat removal with cooling indirectly from the surface of the RPV is outstanding for the DAWS, and a severe-accident-free HTGR can be designed. The newly developed code is applicable to the detailed safety evaluation necessary to future HTGR design.

Evaluation of Tritium Confinement Performance of Alumina and Zirconium for Tritium Production in a High-Temperature Gas-Cooled Reactor for Fusion Reactors

Abstract Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety and recovery efficiency, tritium confinement is an important issue. It is known that alumina has high resistance for gas permeation. In this study, hydrogen permeation experiments in commercial alumina tubes were conducted and hydrogen permeability, diffusivity and solubility were evaluated. By using obtained data, tritium permeation behavior from an Al2O3-coated Li-compound particle was simulated. Additionally, by using literature data for hydrogen behavior in zirconium, an effect of Zr incorporation into an Al2O3 coating on tritium permeation was discussed. It was indicated that the majority of produced tritium was released through the Al2O3 coating above 500 °C. However, it is expected that total tritium leak is suppressed to below 0.67 % of total tritium produced at 500 °C by incorporating Zr fine particles into the inside of Al2O3 coating, assuming tritium pressure inside particle is kept at the plateau pressure of the Zr hydride generation reaction.

Method and results of safety evaluation of the high-temperature engineering test reactor

A modular high-temperature gas-cooled reactor (MHTGR) is expected to be one of the best energy sources in the near future because it can supply high-temperature heat and have high thermal efficiency and sufficient safety features. The safety evaluation of the future MHTGR should be performed based on the experience obtained from the safety evaluation of the High-Temperature Engineering Test Reactor (HTTR). The safety evaluation of the HTTR was performed considering the specific safety design features of the HTGR and is applicable to the future MHTGR. Before the detailed safety evaluation of the future MHTGR, the safety evaluation method and results of the HTTR should be reviewed, and newly established acceptance criteria and methods for selecting evaluation events must be clarified. This paper describes in detail the method and results of the safety evaluation of the HTTR.

Thermal Performance of Intermediate Heat Exchanger during High-Temperature Continuous Operation in HTTR

To use the High-Temperature Gas-cooled Reactor (HTGR) system as a commercial reactor similar to the future HTGR ‘GTHTR300C,’ supplying stable high-temperature heat from the reactor to the heat utilization system such as the hydrogen production system over a prolonged period should be demonstrated. The 50-day high-temperature continuous operation with full power, which is the first long-term operation with a reactor outlet coolant temperature over 900°C, was achieved in March 2010 in the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA). In order to supply stable high-temperature heat from the reactor to the heat utilization system using the HTTR, the heat exchange performance and structure integrity of the intermediate heat exchanger (IHX) should be confirmed. In this paper, the heat exchange performance and structure temperature of the IHX were evaluated with the high-temperature continuous operation data and compared with the designed one or estimated one. As a result, it was confirmed that the IHX can exchange stable high-temperature heat from the primary coolant to the secondary helium, and the IHX structure temperature is under the allowable working temperature in the operation. Moreover, it was confirmed that the heat exchange performance is almost the same as the designed one and that the structure temperature estimated with the analysis code is almost the same as the measured one.

Safety shutdown of the high temperature engineering test reactor during loss of off-site electric power simulation test

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.

Experiments and validation analyses of HTTR on loss of forced cooling under 30% reactor power

In a safety demonstration test involving the loss of both reactor reactivity control and core cooling, the high-temperature engineering test reactor (HTTR) demonstrates spontaneous stabilization of the reactor power. The test and analytical results of tripping one or two out of three gas circulators without reactor scram have already been reported. Moreover, the pre-analytical result of tripping all three gas circulators without reactor scram has been presented. On the other hand, the test and analytical results of tripping all three gas circulators without reactor scram are shown in this paper. About experiments, at an initial reactor power of 30% (9 MW), when all three gas circulators were tripped without reactor scram to reduce the coolant flow rate to zero, the fuel temperature did not show a large increase because the large heat capacity of the graphite core could absorb heat from the fuel in a short period. Moreover, the decay heat could be transferred through the graphite core and the reactor pressure vessel (RPV), emitted by thermal radiation from its outer surface and removed to the active vessel cooling system; therefore, the core at 9 MW was never exposed to the danger of a core melt, and the reactor power was stabilized spontaneously. About analyses, the reactivity performance is important for predicting the converging level of reactor power that affects the fuel temperature during a loss of forced cooling (LOFC) without reactor scram. With regard to thermal hydraulics, the performances of graphite heat conduction in the reactor core and thermal radiation from the RPV surface to the reactor cavity cooling system are crucial for predicting the temperature behavior of the fuel and RPV in the LOFC condition. It was confirmed that reactor kinetics coupled with heat transfer could be applied to reactor safety and accident analysis based on the comparison between the experiments and the analyses.

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.

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.

Study of Tritium Production for Fusion Reactors Using High-Temperature Gas-Cooled Reactors

The performance of a high-temperature gas-cooled reactor as a tritium production device was examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) was assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It was shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 6~10 kg of tritium in a year.