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In ImPACT Tough Robotics Challenge(TRC), field evaluations for several robots with different characteristics have been carried out with several “Tough” tasks aiming to R&D evaluation and social implementation of the robots. In this paper, we describe the development and safety management of the rubble field used in TRC field evaluation, where the safety management is based on international safety standards in order to consider several risk parameters in a systematic way. According to the experiences of the field evaluations, the lessons learned was summarized for further development of the field in future.

Thermal-hydraulics governs mechanical load and environmental condition of structural materials, thus a quantitative understanding of thermal-hydraulic behavior is of real importance to a correct estimate of many kinds of plant aging processes. The current status of thermal-hydraulics for prediction of plant aging is summarized and subjects for future research are discussed.

In the system analyses of a large-break loss-of-coolant accident (LBLOCA) of pressurized water reactors (PWRs), the TRAC-PF1 code predicted an unrealistic depressurization and required much computational time due to the problem of the condensation model. To eliminate the unrealistic depressurization, the TRAC-PF1 code was improved using a simplified condensation model that determined the total condensation rate at cold leg. Through the assessment calculations for CCTF, UPTF and LOFT tests, it was confirmed that the simplified model could eliminate the unrealistic depressurization and reduce the computational time. It was also confirmed that the model could improve the accuracy of the system calculation for the core inlet flow rate and clad temperature as the result of the elimination of the unrealistic depressurization. It has been concluded that the simplified condensation model is useful for the system calculation of the PWR LBLOCA.

A JAERI passive safety reactor (JPSR) system has been developed for reduction of manpower in operation and maintenance and increase of safety. The major features are as follows; an inherent matching nature of core heat generation and heat removal, in-vessel control rod drive mechanism units, once-through steam generators, a large volume pressurizer, passive heat removal system, and passive safety injection system. The passive heat removal system consists of residual heat exchangers, gravity coolant injection pools and air coolers. Steady-state analysis clarified that core residual heat during the whole period from shutdown to refueling phase can be transferred by natural circulation to the gravity coolant injection pool and finally to the atmosphere. In case of a large break loss-of-coolant accident (LOCA), steam is discharged and condensed in the gravity coolant injection pool. It was evaluated that the pool water temperature remains below boiling temperature under the large break LOCA condition. In the course of the design study, probabilistic safety assessment (PSA) was used to clarify the safety features and weak points. The results indicated that loss of offsite power is the dominant initiating event. Based on the analyses, improvements of safety system design were proposed.

The residual heat for the JPSR (JAERI Passive Safety Light Water Reactor) is removed by a natural-circulation of coolant flowing through downcomer. The numerical analysis has been performed taking account of the downcomer being a three-dimensional annulus flow pass with the purposes to confirm the abilities of (1) approximation of three-dimensional thermal fluid flow in downcomer to simple one-dimensional one assumed on the preliminary design of the passive residual heat removal system and (2) achievement of an enough driving-force of the natural circulation to remove the residual heat. The following results were obtained : (1) Flow pattern in downcomer shows remarkable three-dimensional ity(multi-dimensionality)at lower inlet flow rate not to be able to approximate to one-dimensional flow field. However, the temperature distribution does not deviate from uniform one so much even if the multi-dimensional flow such as large vortex arises. (2) It can be expected to obtain the required enough driving-force at a steady state in any case of inlet flow rate where multi-dimensional flow pattern appears. (3)The increase ratio of the driving-force with the time-integrated coolant amount can be estimated as two functional curves in case of higher and other lower inlet flow rates not dependent only on the respective inlet flow rate.

Steady-state and transient critical heat flux (CHF) experiments were performed using triangular-pitch 7-rod assemblies with non-uniform power distributions under the maximum pressure of 15.5 MPa. The onset of steady-state CHF was predicted within an uncertainty of 10% with the KfK correlation using local flow conditions calculated by the subchannel analysis code COBRA-IV-I. However, existing mechanistic CHF models did not agree with the steady-state CHF data. The transient CHFs under the conditions of flow reduction, power increase or flow and power simultaneous variation were predicted with the quasi-steady-state method within approximately the same uncertainty as the steady-state cases. In order to clarify the CHF phenomenon, a real-time neutron radiography technique was used to visualize boiling flow inside a stainless-steel shroud under a pressure of 1.0 MPa. Various two-phase flow patterns and flow behaviors just before and after the onset of CHF were observed by this method.

Thermal-hydraulic behavior in the pressure vessel during the reflood phase of a PWR-LOCA is discussed based on the results from tests with Cylindrical Core and Slab Core Test Facilities, which model a 1100-MWe-class PWR with a scaling ratio of about 1/20. Major findings on core thermal-hydraulics are: (i) substantial water accumulation in the upper part of the core, (ii) radially uniform water accumulation, and (iii) fluid circulation and/or concentration to high power bundles. These multi-dimensional phenomena cause better core cooling in the high power bundle than expected from an evaluation model for licensing based on one-dimensional reflood experiments. After comparison with data from the FLECHT low-flooding tests the substantial water accumulation is considered to be caused by recirculation flow. Based on small scale reflood experiments, effect of local mass flow on heat transfer enhancement in film boiling was correlated and all phenomena except the flow concentration effect were numerically well-simulated by using an analytical model with this correlation. The flow concentration effect on core heat transfer was empirically correlated with local power ratio. It was observed that the core coolability was enhanced by non-uniform fluid mixing between core flow and subcooled water from the hot legs for PWRs with combined-injection ECCS. This is a multi-dimensional phenomena. The heat transfer enhancement was also well-simulated with the above-mentioned model.

A conceptual design of a double-flat-core type high conversion pressurized water reactor (HCPWR) has been developed at the Japan Atomic Energy Research Institute to improve fuel utilization. In thermal-hydraulic feasibility studies of the proposed concept, major six events such as large and small break LOCAs have been analyzed with the best estimate code REFLA/TRAC. This paper describes the results of control rod ejection accident analyses as a part of the series of analyses. The calculations were performed according to the licensing analysis on current PWRs. The calculated results showed that the proposed concept had larger safety margines than current PWRs. Therefore, the proposed concept is considered to be feasible from a view point of safety analysis of control rod cruster ejection accident. It is seen that the large safety margine of the proposed concept is originated from smaller excess reactivity, smaller maximum linear heat generation rate and smaller rod worth per a control rod than those of current PWRs.

Research and developmental work on the high-conversion light water reactor (HCLWR) at JAERI aims at attractive design in terms of energy source and economy to ensure a safety level equal to or higher than that of current LWRs. In thermal hydraulics, efforts are being devoted to understand the phenomena in a tight lattice core and to develop analytical methods for optimum design. In critical heat flux (CHF) experiments, several existing CHF correlations were evaluated. It was found that the KfK correlation can be applied to the triangular tight lattice rod array used for the experiment reported here. In the hydraulic experiment the core pressure drop was investigated with existing correlations in various types of core geometries. It was found that Rehme's correlation for spiral wire and integral spiral rib can be applied and that the drag coefficient of grid spacers is noticeably affected by the height of spacers. On the basis of these experiments, the thermal-hydraulic characteristic was investigated for the double-flat-core HCLWR, which potentially has advantages such as negative void reactivity coefficient, high conversion ratio, and high burnup. It was found that the double-flat-core HCLWR is highly feasible, having enough margin in CHF, a core pressure drop comparable to current PWRs, and a fairly low peak clad temperature in LOCA reflood as calculated by REFLA code.

The scale effects of a flow path under countercurrent flow limitation (CCFL) (air/water or steam/saturated water) in a horizontal tube connected to an inclined riser have been studied. The studied geometry simulates that of a PWR hot leg. An analytical model with a two-fluid model was developed based on flow observation results in small scale experiments and then assessed with various scale experiments under various pressures to evaluate the scale effects. The assessments with small scale experiments revealed that the region controlling the flow limitation was shifted from the horizontal tube near the bend to the inclined riser part as the length of inclined riser increased. The degree of the shift became weaker for larger scale experiments under the assumption that the ratio of interfacial friction factor to wall-to-gas friction factor was maintained to be the same as that for small scale experiments. The degree of the shift was not affected by the change of pressure ( 0.3 MPa → 1.5 MPa ).

Reflood phenomena of 2-loop PWR with upper-plenum-injection (UPI) type ECCS was studied by examining CCTF experimental data. The major conclusions about the effects of asymmetric UPI and single low-pressure-coolant-injection (LPCI) pump failure on reflood phenomena are as follows: 1. Single failure assumption of LPCI pump is more conservative (worse core cooling) than no failure assumption of LPCI pump under UPI condition. This is consistent with current guide line for reactor safety evaluation. 2. The asymmetric UPI promoted a radial ununiformity of the core cooling, although even symmetric UPI lead to the radial ununiformity of the core cooling. 3. The asymmetric UPI did not give a local degradation of core cooling in comparison with symmetric UPI but gave an enhancement of core cooling in the upper part of core after the initiation of water fall back. 4. The enhancement of core cooling is considered to be caused by increased upward fluid flow due to water fall back and water fall back itself, which are...

The effects of the ECC water injection rate on reflooding behavior during a PWR-LOCA have been investigated using the experimental data from a large scale reflood test facility. As a general result, the higher ECC water injection rate was found to result in the higher core flooding rate, and hence, the better core cooling. However, when the LPCI injection rate was higher than a certain value, which was enough to keep the downcomer almost filled, the core flooding rate was found to become an almost identical value being regardless of the LPCI injection rate. Also found was the effect of the Acc injection rate and duration influenced the core cooling behavior through the whole transient including the LPCI period. These phenomena have been analyzed and clarified quantitatively in this paper. It was also found that the Acc injection rate and duration effects on reflooding behavior observed in the large scale facility were almost the same as those already observed in the small scale facility.

Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWRLOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047 K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20 s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo...

The over-all core hydrodynamics during the reflood phase is quantitatively discussed. The following model was proposed for the core hydrodynamics: (1) When the liquid mass flux and the steam mass flux exist at the quench front, dispersed flow appears. (2) The void fraction downstream of the quench front gradually decreases with time. (3) Finally, the void fraction approaches the value predicted by the modified Cunningham & Yeh and the modified Lockhart-Martinelli correlations within about ±30% error. (4) Even after the concerning region is quenched, the void fraction can be predicted with the same correlations. Both correlations have been recently developed.

Emergency core cooling (ECC) mater is carried up to the upper plenum and falls down again into the core during the reflood phase in PWR-LOCA. Therefore the quench front also propagates downward from the top of the core. The effect of upward steam flow rate on the top-down quench propagation was experimentally investigated. It was found that top-down quench velocity was delayed by upward steam flow. This effect is more significant when rod surface temperature is low and the falling water flow rate is small. The effect of the flow rate and the rod temperature on the quench velocity was correlated based on the experimental results under the conditions of atmospheric pressure, saturation temperature for water and steam, rod surface temperature of 350–600°C, down-ward water velocity of 0.01–0.1 m/s and upward steam velocity of 0–20 m/s.

A simple model was developed to evaluate the effect of the radial power profile on the thermo-hydraulics in the core during reflood phase of loss-of-coolant accident in PWRs. It was incorporated into REFLA code, a reflood analysis code. The model assumes that the fluid mixes completely among subchannels and additionally that the fuel temperature is not affected each other among fuels. This model is featuring no three-dimensional treatment for the evaluation on effect of the radial power profile. The model improved the predictability of clad temperature and gave a better prediction than a method which is generally used in the safety analysis. The model predicted a poorer core cooling in high radial power region than data. The error of the prediction increased with the increase in radial power ratio. These indicates the conservativeness of the model on the maximum clad temperature prediction. The reason of the conservativeness is due to the neglection of the heat transfer enhancement by two-dimensional effe...

The conservatism of the current safety analysis was assessed by comparing the predicted results with cylindrical core test facility (CCTF) test results performed at Japan Atomic Energy Research Institute (JAERI). The WREM code was selected for the assessment. The overall conservatism of the WREM code on the peak clad temperature prediction was confirmed against CCTF EM test which simulated the typical initial and boundary conditions in the safety evaluation analysis. The WREM code predicted the reasonable core boundary conditions and the conservatism of the code came mainly from the core calculation. The conservatism of the WREM code against CCTF data could be attributed to the following three points: 1. (i) no horizontal mixing assumption between subchannels at each elevation, 2. (ii) no modeling on heat transfer enhancement caused by the radial core power profile, 3. (iii) usage of conservative heat transfer correlations in the code.