Yukio Sudo

Present status of the High Temperature Engineering Test Reactor (HTTR)

Abstract In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis. The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR. The installation permit of the HTTR was issued by the Government in November 1990.

Evaluation of core thermal and hydraulic characteristics of HTTR

Abstract 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% and 1492°C, respectively.

Experimental study on the effects of channel gap size on mixed convection heat transfer characteristics in vertical rectangular channels heated from both sides

Abstract The effects of channel gap size on mixed forced and free convective heat transfer characteristics were experimentally investigated for water flowing near atmospheric pressure in a 750 mm long and 50 mm wide channel heated from both sides. The channel gap sizes investigated were 2.5, 6, 18 and 50 mm. Experiments were carried out for both aiding and opposing forced convective flows with a Reynolds number Re x of 4 × 10 to 6 × 10 6 and a Grashof number Gr x of 2 × 10 4 to 6 × 10 11 , where the distance x from the inlet of the channel is adopted as the characteristic length in Re x and Gr x . As for the results, the following were revealed for the parameters ranges investigated in this study. 1. (1) When the dimensionless parameter, Gr x / Re x 21 8 Pr 1 2 is less than 10 −4 , the flow shows the nature of forced convective heat transfer for a channel with any channel gap size in both aiding and opposing flows. 2. (2) When Gr x / Re x 21 8 Pr 1 2 is larger than 10 −2 , the flow shows the nature of free convective heat transfer for a channel with any channel gap size in both aiding and opposing flows. 3. (3) When Gr x / Re x 21 8 Pr 1 2 is between 10 −4 and 10 −2 for the channel with a channel gap size equal to or larger than 6 mm, the heat transfer coefficients in both aiding and opposing flows become, on the average, higher than those predicted by the previous correlations for either the pure turbulent forced convection or the pure free convection, and can be expressed in simple forms with a combination of Gr x / Re x 21 8 Pr 1 2 and the previous correlation for either the pure turbulent forced convection or the free convection along a flat plate. 4. (4) When Gr x / Re x 21 8 { Pr 1 2 is between 10 −4 and 10 −2 for the channel with a channel gap size of 2.5 mm, the heat transfer coefficients in both aiding and opposing flows also become, on the average, higher than those predicted by the previous correlations for either the pure turbulent forced convection or the pure free convection. This is considered to be because the acceleration of the main flow originated by the development of the boundary layer in a narrow rectangular channel promotes the heat transfer.

Effects of Radial Core Power Profile on Core Thermo-Hydraulic Behavior during Reflood Phase in PWR-LOCAs

An investigation of the effects of the radial core power profile on the thermo-hydraulic behavior during the reflood phase in a PWR-LOCA has been conducted with the Slab Core Test Facility (SCTF). Since the power in an actual PWR is lower in the peripheral bundles than in the central bundles, the so called chimney effects due to radial core power profile are expected to improve the cooling of the higher power bundles. The SCTF simulates a full radius slab section of a PWR and therefore the effects of radial core power profile can be investigated. The revealed results of four forced-feed reflood tests in the SCTF are; (1) even with different radial core power profiles, flat distribution of the collapsed water level in the core are obtained for each test; (2) in the highest power bundle under the same total core power, steeper radial power profile gives higher heat transfer coefficient; and (3) redistribution of flow or cross flow between bundles is considered to be a major reason for the results described ...

Depressorization accident analysis for the FTTR by the TAC-NC

The two-dimensional thermal analysis code TAC-NC is modified from the analytical code TAC-2D in order to calculate temperature transients in the case of loss of forced cooling accidents of the HTTR (High Temperature engineering Test Reactor) such as a depressurization accident. The TAC-NC code includes a special function to calculate heat transfer by natural convection between hotter and colder regions in the pressure vessel during the depressurization accident.

Design and safety consideration in the High-Temperature engineering Test Reactor (HTTR)

The High Temperature engineering Test Reactor (HTTR) is a test reactor in Japan with thermal output of 30MW and reactor outlet coolant temperature of 950 °C at a high temperature test operation. This report describes features of the HTTR, emphases being laid on the safety design.

Critical heat flux at high velocity channel flow with high subcooling

Abstract A quantitative analysis of critical heat flux (CHF) in heated channels under high mass flux with high subcooling was successfully carried out by applying a new flow model to the existing CHF model of a macro-water-sublayer on the heated wall and steam blankets over it. The CHF correlation proposed could correctly predict the existing experimental data for circular tubes of 0.33–4 mm in diameter with mass flux of 124–90 000 kg (m 2 s) −1 and inlet water subcooling of 35–210 K at 0.1–7.1 MPa, resulting in CHF of 4.2–224 MW m −2 , and for rectangular channels of 3–20 mm gap with a mass flux of 940–27 000 kg (m 2 s) −1 and inlet water subcooling of 13–166 K at 0.1–3.0 MPa, resulting in CHF of 2.0–62 MW m −2 . An error of the CHF correlation has also been estimated.

Downcomer Effective Water Head during Reflood in Postulated PWR LOCA

A study was conducted of the downcomer effective water head which is the only driving force to supply emergency core coolant into core during reflood phase in a PWR loss-of-coolant accident. With a full height downcomer simulator, effective water head experiments were carried out under 1 atm to investigate the applicability of the correlation for void fraction for evaluating the effective water head as well as to investigate the effect of the scale factor. As the results, the effect of the scale factor (the gap of the downcomer) was revealed to be significant, that is, the smaller gap gives the smaller effective water head. From the comparison of predictions based on the correlation for void fraction with the experimental results, it was revealed that (1) for a slow effective water head change expected in the reflood phase, the used correlation gives a good prediction for the experimental results and that (2) for a rapid change of effective water head, however, more elaborate investigation is needed. It w...

Present status and prospect of microbeams at TIARA

JAERI had made a new research program so-called the advanced radiation technology project in 1987, and conducted microbeam studies in the fields of semiconductor devices, analytical technique and biological application using different type of microbeam systems installed at Takasaki Ion Accelerators for Advanced Radiation Application (TIARA) which is an ion beam irradiation facility for making the promotion of the project. The present status of these microbeam studies is summarized, and the future prospects are discussed.

Design and evaluation methods for a water cooling panel system for decay heat removal from a high‐temperature gas‐cooled reactor

An experiment was performed to simulate a water cooling panel system for decay heat removal from a high-temperature gas-cooled reactor (HTGR) and to investigate the performance of decay heat removal and the temperature distribution for components of the system. The experimental apparatus is composed of a pressure vessel 1 m in diameter and 3 m in height, containing heaters with a maximum heating rate of 100 kW which simulates the decay heat of the reactor core and cooling panels surrounding the pressure vessel. The analytical code THANPACST2 was applied to analyze the experimental data and to investigate the validity of the analytical method and model proposed. Under conditions using helium gas at a pressure of 0.73 MPa and temperature of 210°C in the pressure vessel, temperatures of the pressure vessel were well estimated to within differences of -29 to +37°C compared to the experimental data. The analyses indicate that the heat removed by the cooling panel is 11.4% less than the experimental value and the heat transferred by thermal radiation is 74.4% of the total heating value. It was also found/that the lower head of the pressure vessel is effectively cooled by natural convection through the flow channels at the upper and the lower edges of the skirt-type support of the pressure vessel.