Loss Of Forced Cooling Research Articles

Improved multiphysics model of the High Temperature Engineering Test Reactor for the simulation of loss-of-forced-cooling experiments

We present a multiphysics model of the High Temperature Engineering Test Reactor for comparison with past and predict future loss-of-forced-cooling (LOFC) experiments. The approach selected combines (1) 3-D full-core superhomogenization-corrected neutronics, (2) 3-D full-core homogenized or semi-heterogeneous heat transfer (macroscale), (3) 2-D axisymmetric fuel rod heat transfer (pin-scale), and (4) 1-D thermal-hydraulics channels. Although large uncertainties remain, the time and magnitude of the first power peak after re-criticality is predicted within 1.5 h and 175 kW, respectively. The novelty of our work includes (1) a new macroscale/pin-scale heat transfer coupling approach relying on gap conductance to drastically speed up numerical convergence by two orders of magnitude, (2) determination of a radial effective thermal conductivity, reproducing the semi-heterogeneous re-criticality time within one hour using a homogenized macroscale model, and (3) a preliminary study of the reactor’s early behavior following a LOFC event, enabling further assessment of numerical models against fission power measurements.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

Loss of Core Cooling Test With One Cooling Line Inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured temperatures of RPV and the air outside of biological shielding concrete. The temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 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) demonstrated spontaneous stabilization of the reactor power. A test where all three gas circulators were tripped at 30% reactor power (9MW) without a reactor scram, called a loss-of-forced-cooling (LOFC) test, was analyzed, and the analytical results from a reactor kinetics code was reported. After that, the Japan Atomic Energy Agency (JAEA) acquired a large supercomputer and the STAR-CCM+ software developed by CD-adapco, which can be used to create a three-dimensional model of the HTTR to analyze steady-state and transient conditions. In this paper, a model of the three-dimensional thermal-hydraulics inside the reactor pressure vessel (RPV) during the LOFC test at 30% reactor power (9MW) is described and detailed analyses are performed using STAR-CCM+. With this tool, the effects of natural and forced convection on thermal-phenomena in the HTTR can be understood quantitatively. The finding is as follows: the downstream of forced convection caused by the helium purification system (HPS) pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection due to the HPS has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. And the effect can be neglected because the difference of the maximum velocity of the helium between the cases where the HPS is active and inactive, is very little over the course of LOFC test. The proposed three-dimensional analytical model will be useful for three-dimensional accident analyses.

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.

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.

Thermal-hydraulic design and transient evaluation of a small long-life HTR

Small long-life high temperature gas-cooled reactors (HTRs) may provide electricity or heat for remote areas or industrial users in developed and/or developing countries. Moreover, small HTRs have advantages over large nuclear reactors of demonstrated inherent safety, transportability, modular construction, and flexible site selection. This paper presents the thermal-hydraulic evaluations of the U-Battery, which is a small, long-life and block-type HTR using the DALTON/THERMIX code system. The thermal-hydraulic characteristics of a cylindrical design and an annular design of the U-Battery were evaluated for loss of forced-cooling (LOFC) incidents including depressurized LOFC (DLOFC) and pressurized LOFC (PLOFC) incidents. The calculations show that the stronger natural circulation during the PLOFC makes the reactor core cool faster than during the DLOFC, flattens the radial solid temperature distribution, and transfers more heat from the hot regions (bottom and center of the reactor core) to cold regions (top and periphery of the reactor core). Although the natural circulation in the reactor core is so weak that it is neglected during the DLOFC, the decay heat is removed passively by conduction without any violation of the temperature limits for the 20MWth U-Battery. The comparisons of the cylindrical and annular reactor core configurations show that the latter is a better design with a lower maximum core temperature during the LOFC because of the central graphite reflector. Moreover, it is possible to adopt the current reactor configuration when the thermal power of the U-Battery increases from 20MWth to 40MWth.

Validation of the DALTON-THERMIX Code System with Transient Analyses of the HTR-10 and Application to the PBMR

The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5.The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature.For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor–10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system.It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.