Achievement of Reactor-Outlet Coolant Temperature of 950°C in HTTR

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30 MW and reactor-outlet coolant temperature of 950°C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950°C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950°C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.     

Safety Analysis of Abnormal Reactivity Events in HTTR

Abstract

To confirm the safety of the High Temperature Engineering Test Reactor (HTTR) facility which is being constructed as the first high temperature gas cooled reactor in Japan, the representative abnormal reactivity events assumed in the safety analysis of the HTTR were analyzed. The HTTR is a graphite moderated and He-gas-cooled reactor with thermal power of 30 MW, inlet coolant temperature of 395°C and outlet coolant temperature of 950°C.

This report presents the analytical results of two representative events, “Abnormal control rod withdrawal from a subcritical condition” and “Abnormal control rod withdrawal during the full power operation”, showing that the safety of the HTTR is secured in conformity with the unique features of the HTTR with respect to the maximum fuel temperature, which is a key factor for the safety criteria.

The results of the safety analysis could demonstrate the safety of the HTTR facility with respect to abnormal reactivity events postulated in the HTTR, showing that the maximum fuel temperature is lower than the limit of the maximum fuel temperature of 1,600°C.     

Evaluation of Hot Spot Factors for Thermal and Hydraulic Design of HTTR

Abstract

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.     

Integrity assessment of the high temperature engineering test reactor (HTTR) control rod at very high temperatures

The high temperature engineering test reactor (HTTR) is the first high temperature gas-cooled reactor (HTGR) in Japan with a reactor outlet coolant temperature of 950°C at high temperature test operation. The HTTR contains 16 pairs of control rods for which Alloy 800H is chosen of the metallic parts. Because the maximum temperature of the control rods reaches about 900°C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Under the guideline, temperature and stress analysis was conducted, and it is confirmed that the target life of the control rods of 5 years can be achieved.     

Numerical simulation of three-dimensional thermal-hydraulic behavior for HTTR (High Temperature Engineering Test Reactor)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are now in progress in order to verify the inherent safety features and to improve safety designing and analysis technologies for future HTGR (high temperature gas-cooled reactor). Coolant flow reduction test is one of the safety demonstration tests for the purpose of demonstration of inherent HTGR safety features in the case that coolant flow is reduced by tripping of helium gas circulators. If reactor core element temperature and core internal structure temperature during abnormal events are estimated by numerical simulation with high-accuracy, developed numerical simulation method can be applied to future HTGR designing efficiently. In the present research, three-dimensional in- and ex-vessel thermal-hydraulic calculations for the HTTR are performed with a commercially available thermal-hydraulic analysis code “STAR-CD^®” with finite volume method. The calculations are performed for normal operation and coolant flow reduction tests of the HTTR. Then calculated temperatures are compared with measured ones obtained in normal operation and coolant flow reduction test. As the result, calculated temperatures are good agreement with measured ones in normal operation and coolant flow reduction test.     

Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents

The high temperature engineering test reactor (HTTR) being constructed by the Japan Atomic Energy Research Institute is a graphite-moderated, helium-cooled reactor with an outlet gas temperature of 950 °C.Two independent vessel cooling systems (VCSs) of the HTTR cool the reactor core indirectly during depressurized and pressurized accidents so that no forced direct cooling of the reactor core is necessary. Each VCS consists of a water cooling loop and cooling panels around the reactor pressure vessel (RPV). The cooling panels, kept below 90 °C, cool the RPV by radiation and natural convection and remove the decay heat from the reactor core during these accidents.This paper describes the design details and safety roles of the VCSs of the HTTR during depressurized and pressurized accidents. Safety analyses prove that the indirect core cooling by the VCSs and the inherent safety features of the reactor core prevent a temperature increase of the reactor fuel and fission product release from the reactor core during these conditions. Furthermore, it is confirmed that even if VCS failure is assumed during these accidents, the reactor core and RPV can remain in a safe state.     

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data.     

Reactor core design of Gas Turbine High Temperature Reactor 300

Japan Atomic Energy Research Institute (JAERI) has been designing Japan’s original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h.

This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.     

Reactor technology development under the HTTR project

JAERI is developing HTR technology, hydrogen production technology, and system integration technology under the HTTR Project. The HTTR is the Japanese first HTR with a 30-MW thermal power. The first criticality of the HTTR was achieved in 1998, and the full-power operation at an outlet coolant temperature of 850°C was attained in 2001. The outlet coolant temperature was reached to 950°C in 2004. A seven-year program on the gas turbine HTR was launched in 2001. The program consists of the design of a GTHTR300 plant and R&D on a closed-cycle helium gas turbine system for the GTHTR300. It is designed to have a 600-MW thermal power at an outlet coolant temperature of 850°C and a 275-MW electric power. The objectives of the program are to establish a feasible plant design and to demonstrate key technologies for the helium gas turbine. The GTHTR300 design will demonstrate competitive economy and high degree of safety. It will also provide technology basis of VHTRs for power generation, hydrogen production, and cogeneration.     

Inspection techniques for primary pressurized water cooler tubes in the high temperature engineering test reactor

A primary pressurized water cooler (PPWC) with 136 inverse-U-tubes is installed in the primary cooling system of the high temperature engineering test reactor (HTTR). The HTTR is the first high temperature gas-cooled reactor in Japan with an outlet gas temperature of 950 °C and thermal power of 30 MW. The heat transfer tubes form the reactor pressure boundary of the primary coolant. Inspection techniques for the tubes should be established to carry out the in-service inspection efficiently. An automatic inspection system for the tubes uses probes for eddy current testing and ultrasonic testing. Defect detecting characteristics of the inspection probes and the application of the automatic inspection system to nondestructive test of the tubes were examined by a mockup test utilizing artificially degraded tubes. The automatic inspection system could smoothly insert and withdraw the probe at its moving velocity in the fixed positions of the defected tube. Nondestructive test of the tubes using the automatic inspection system was conducted during reactor shutdown period of the HTTR after test operation of about 55% of the full power. Through the nondestructive test, it was concluded that there was no defect on the outer surface of the heat transfer tubes of the PPWC inspected.     

Coal gasification with heat from high temperature reactors: Objectives and status of the project “Prototype Plant for Nuclear Process Heat (PNP)”

It is by now fairly widely known that the high temperature reactor (HTR) is a unique nuclear energy source which can supply heat at temperatures up to 1000°C for application in chemical processes, for which previously exclusively combustion heat sources have been used. With the HTR, it is possible to apply nuclear energy not only for electrical power production but also for the synthesis of liquid or gaseous energy carriers. Nuclear coal gasification appears most promising as the first step for the demonstration and industrial application of nuclear process heat technology in the Federal Republic of Germany. Reactor manufacturers and coal mining companies in co-operation with the Nuclear Research Center at Jülich established a joint project in 1975 for the development of an HTR with a coolant outlet temperature of 950°C, for the development and testing of nuclear coal gasification, for the detailed engineering of a prototype plant consisting of an HTR and gasification plant and finally for the construction and operation of this prototype plant for nuclear process heat (PNP). This contribution describes the status of the PNP-Project and the scope for future development.     

Validation of the nuclear design code system for the HTTR using the criticality assembly VHTRC

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.     

Simulation tests for temperature mixing in a core bottom model of the HTR-Module

Interatom and Siemens are developing a helium-cooled Modular High Temperature Reactor. Under nominal operating conditions temperature differences of up to 120°C will occur in the 700°C hot helium flow leaving the core. In addition, cold gas leakages into the hot gas header can produce even higher temperature differences in the coolant flow. At the outlet of the reactor only a very low temperature difference of maximum ±15°C is allowed in order to avoid damages at the heat exchanging components due to alternating thermal loads. Since it is not possible to calculate the complex flow behaviour, experimental investigations of the temperature mixing in the core bottom had to be carried out in order to guarantee the necessary reduction of temperature differences in the helium. The presented air simulation tests in a 1:2.9 scaled plexiglass model of the core bottom showed an extremely high mixing rate of the hot gas header and the hot gas duct of the reactor. The temperature mixing of the simulated coolant flow as well as the leakage flows was larger than 95%. Transfered to reactor conditions this means a temperature difference of only ±3°C for the main flow at a quite reasonable pressure drop. For the cold gas leakages temperature differences in the hot gas up to 400°C proved to be permissible. The results of the simulation experiments in the Aerodynamic Test Facility of Interatom permitted to design a shorter bottom reflector of the core.     

Structural design of high temperature metallic components

In the high temperature engineering test reactor (HTTR), even at normal operation the service temperatures of class 1 metallic components reach temperatures above 900 °C when exposed to primary helium coolant of 950 °C. For these components, Hastelloy XR, which is the improved version of Hastelloy X, was developed and used for high temperature application.Some of the high temperature materials and their service temperatures, including Hastelloy XR, used for the class 1 and reactor internal metallic components of the HTTR are very well beyond the well-established Japanese elevated temperature structural design guideline. Moreover, at very high temperatures, where creep deformation is significant, the component design based on elastic analysis is impossible. Therefore, many research works on structural mechanics behavior were carried out to establish a high temperature structural design guideline and creep analysis methods. This paper reviews structural design of the high temperature components for the HTTR made of Hastelloy XR, 2 1/4Cr–1Mo steel, austenitic stainless steels SUS321TB and SUS316, and 1Cr–0.5Mo–V steel.     

Design of a supercritical water-cooled reactor with a three-pass core arrangement

The Supercritical Water-cooled Reactor (SCWR) is one of the six concepts of the Generation IV International Forum. In Europe, investigations have been integrated into a joint research project, called High Performance Light Water Reactor (HPLWR). Due to the higher heat up within the core and a higher outlet temperature, a significant increase in turbine power and thermal efficiency of the plant can be expected.Besides the higher pressure and higher steam temperature, the design concept of this type of reactor differs significantly from a conventional LWR by a different core concept. In order to achieve the high outlet temperature of over 500 °C, a core with a three-step heat up and intermediate mixing is proposed to keep local cladding temperatures within today's material limits. A design for the reactor pressure vessel (RPV) and the internals has been worked out to incorporate a core arrangement with three passes. All components have been dimensioned following the safety standards of the nuclear safety standards commission in Germany. Additionally, a fuel assembly cluster with head and foot piece has been developed to facilitate the complex flow path for the multi-pass concept. The design of the internals and of the RPV is verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Furthermore, the reactor design ensures that the total coolant flow path remains closed against leakage of colder moderator water even in case of large thermal expansions of the components. The design of the RPV and internals is now available for detailed analyses of the core and the reactor.     

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States.     

Fracture mechanical properties and neutron irradiation effects of fuel compacts for the HTTR

To simulate the nuclear fuel for High Temperature Engineering Testing Reactor (HTTR), fuel compact models using SiC-kernel coated particles instead of UO₂-kernel coated particles were prepared under the same conditions as those for the real fuel compact. The mechanical and fracture mechanics properties were studied at room temperature. The thermal shock resistance and fracture toughness for thermal stresses of the fuel compact were experimentally assessed by means of arc discharge heating applied at a central area of the disk specimens. These model specimens were then neutron irradiated in the Japan Material Testing Reactor (JMTR) for fluences up to 1.7 × 10²¹n/cm² (E ·> 29 fJ) at 900°C ± 50°C. The effects of irradiation on a series of fracture mechanical properties were evaluated and compared with the cases of graphite IG-110 used as the core materials in the HTTR.     

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).     

Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 °C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 °C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 °C, respectively, at full power operation and at the scram from the operation.     

European supercritical water cooled reactor

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 °C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.