Procedure to prevent temperature rise of primary upper shielding in high temperature engineering test reactor (HTTR)

In a non-nuclear heat up test of the high temperature engineering test reactor (HTTR), which was conducted in February 1997, the temperature of primary upper shielding as well as helium gas temperature inside standpipes of the reactor became much higher than expected. Because it was estimated that these temperatures should exceed design values at full power operation of the HTTR, countermeasures were considered and taken to prevent the temperature rise. After applying two countermeasure steps, confirmation tests to demonstrate the effect of the measures were performed. The test results and extrapolation by temperature analysis finally showed that the countermeasures are appropriate, and the design temperatures should be observed at full power operation.     

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.     

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.     

Evaluation of aseismic integrity in the HTTR core-bottom structure V. On the static and dynamic behavior of graphitic HTTR key-keyway structures

The graphite components in high temperature gas-cooled reactors are connected to each other through a key-keyway structure that has gaps between the key and the keyway to accomodate thermal expansion. Because a dynamic load concentrates on the key-keyway structure during earthquakes, it is considered to be a crucial element for assessing the integrity of the graphite components. A combination of experiments and analyses was employed to investigate the dynamic behavior of the key-keyway structure, i.e. the equivalent stiffness associated with vibrational characteristics of the graphite components and the stress distribution under dynamic loading. The experiments were performed using a graphite scale model and a dynamic photo-elastic method. The analysis was carried out using the finite element method (FEM) code Abaqus, taking account of the contact between the key and the keyway. The following conclusions were derived. (1) The equivalent stiffness of the key-keyway structure shows nonlinearity, owing to the contact deformation. (2) The equivalent stiffness evaluated by the FEM analysis, taking account of the non-inear contact deformation, is applicable for predicting the vibrational characteristics of ky-keyway structure. (3) The stress concentration under dynamic loading is lower than or nearly equal to that under static loading. The maximum stress concentration of the seismic load can be sufficiently evaluated under static loading conditions.     

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).     

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.     

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.     

Leak-tightness characteristics concerning the containment structures of the HTTR

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area.The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values.The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.     

Study on control characteristics for HTTR hydrogen production system with mock-up test facility: System controllability test for fluctuation of chemical reaction

The Japan Atomic Energy Research Institute has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High-Temperature Engineering Test Reactor (HTTR). Prior to the coupling of a hydrogen production plant with the HTTR, simulation tests with a mock-up test facility of the HTTR hydrogen production system (HTTR-H2) is underway. The test facility is a 1/30-scale of the HTTR-H2 and simulates key components downstream from an intermediate heat exchanger of the HTTR. The main objective of the simulation tests is the establishment and demonstration of control technology, focusing on the mitigation of a thermal disturbance to the reactor by a steam generator (SG) and on the controllability of the pressure difference between the helium and process gases at the reaction tube in a steam reformer (SR). It was confirmed that the fluctuation of the outlet helium gas temperature at the SG and the pressure difference in the SR can be controlled within the allowable range for the HTTR-H2 in the case of the system controllability test for the fluctuation of chemical reaction. In addition, a dynamic simulation code for the HTTR-H2 was verified with the obtained test data.     

高庙子膨润土在电子辐照和高温老化后热稳定性研究

为研究膨润土的长期稳定性,以高庙子天然钙基膨润土及其改性钠基膨润土为研究对象,利用热重一差示扫描量热（TG-DSC）方法对不同累积剂量的电子辐照和不同温度、不同热作用时间下的老化样品进行了分析。结果表明：高温作用引起的膨润土吸热峰峰值温度和起始温度的变化比电子辐照的大;钠基土比钙基土有更好的耐电子辐照和高温老化能力。在电子辐照／高温老化条件下,改性高庙子钠基膨润土的结构热稳定性比其钙基膨润土好。     

Present status of energy in Japan and HTTR project

An amount of primary energy supply in Japan is increasing year by year. Much energy such as oil, coal and natural gas is imported so that the self-sufficiency ratio in Japan is only 20% even if including nuclear energy. An amount of energy consumption is also increasing especially in commercial and resident sector and transport sector. As a result, a large amount of greenhouse gas was emitted into the environment. Nuclear energy plays the important role in energy supply in Japan.Japan Atomic Energy Research Institute (JAERI) has been carried out research and development of a hydrogen production system using a high temperature gas cooled reactor (HTGR). The HTTR project aims at the establishment of the HTGR hydrogen production system. Reactor technology of the HTGR, hydrogen production technology with thermochemical water splitting process and system integration technology between the HTGR and a hydrogen production plant are developed in the HTTR project.     

The temperature dependence of irradiation creep

The stress induced absorption mechanism (SIPA) of irradiation creep will be discussed by developing a new application of rate theory which emphasizes particularly the significance of the vacancy loops formed from the displacement cascades during fast neutron irradiation. A simple analytic result for the expected creep strain rate and various related numerical results will be discussed. The relation between irradiation creep and void swelling will be emphasized, particularly in relation to the significance of the vacancy emission processes from the vacancy loops.     

The theory of phase stability under irradiation

Irradiation-alteration of phase stability is examined from the thermodynamic and kinetic viewpoints. It is shown first that the flux of irradiating particles cannot be treated as a thermodynamic variable, so that an irradiated system is inherently in a driven irreversible state. A typical breeder reactor cladding or fusion first wall irradiation is seen to insert huge (> 10⁷ J/mol) amounts of energy into the material; this energy has the potential to do large amounts of work, given the existence of a coupling mechanism. The various coupling mechanisms are reviewed, and estimates made of their effectiveness. Irradiation-induced solute segregation, disordering, and the chemical vacancy effect are shown able to give changes in stability of up to 10³ J/mol. By contrast, the excess vacancy concentration is unlikely to increase the free energy of the irradiated metal by as much as 1 J/mol. The various mechanisms have the potential to give different kinds of alterations of phase stability, i.e., shift phase boundaries, produce unknown phases, or enhance diffusion. These are discussed and the potentials and limitations of each mechanism are enumerated.     

Phenomenological description of the stability of binary alloys under irradiation

L. Ya. Karpova Scientific-Research Institute of Physical Chemistry. Translated from Atomnaya énergiya, Vol. 76, No. 2, pp. 98–103, February, 1994.     

Thermo-mechanical test rig for experimental evaluation of thermal conductivity of ceramic pebble beds

The experimental determination of mechanical and thermal properties of ceramic pebble beds, such as the lithium orthosilicate or lithium metatitanate, is a key issue in the framework of fusion power technology, for the reason that they are possible candidates in the design of breeder blankets.The paper deals with an experimental method for the evaluation of the thermal conductivity of ceramic pebble beds versus the temperature and compressive strain, based on a steady state heat flux through a material (alumina) of known conductivity. The alumina thermal conductivity is determined by means of the hot wire method. To assess the experimental method, a thermo-mechanical characterization of alumina pebble beds (a material largely available), having different diameters, considering a wide range of temperatures and compression forces has been carried out.Moreover preliminary tests have been performed on lithium orthosilicate and lithium metatitanate pebble beds.     

Studies on irradiation stability of polystyrene by NMR

The irradiation stability of polystyrene (PS) was studied by ^13C and ^1H NMR spectra, Nuclear Overhauser Relaxation (NOE) and ^13C NMR spin-lattice relaxation time (T1). The results indicate that ^13C and ^1H NMR chemical shifts, NOE and T1 were almost invariant with the increase of irradiation dose. This shows that polystyrene is particularly stable within 2.5 kGy doses and the mechanism of its stability is discussed.     

Fuel irradiation of the first batches produced for the Chinese HTR-10

An irradiation test of four spherical fuel elements (SFE) had been performed in the Russian reactor IVV-2M. The elements were sampled randomly from the first and second product batches which were manufactured for the 10 MW high-temperature gas-cooled test reactor (HTR-10). The maximum burnup of the irradiated fuel elements reached 107,000 MWd/tU and the maximum fast neutron fluence was 1.31 × 10²⁵ m⁻². The release-to-birth rate ratio (R/B) did not increase significantly during irradiation. However, an in-pile heating-up test of element SFE 7 in Capsule 5 led to a failure of approximately 6% of the coated particles. After the test it was estimated that the fuel temperature had very likely been much higher than the intended 1600 °C.