Takamichi Iwamura

Conceptual designing of reduced-moderation water reactor with heavy water coolant

Abstract The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06–1.11.

Transport Mechanism of Thermohydraulic Instability in Natural Circulation Boiling Water Reactors during Startup

This paper presents experimental study on transport mechanism of thermohydraulic instability, which may occur in natural circulation boiling water reactor during startup. The research was carried out using a natural circulation experimental loop featuring twin parallel boiling channels with chimney assembly. The experiments were performed with the pressure range of 0.1 to 0.7MPa and maximum heat flux of 577kW/m2. The objective of the study is to formulate thermohydraulic stability maps required for determining rational startup procedure of the reactor, in which the instability could be prevented. The study clarified that the flow modes during startup consist of the following sequence: (1) single-phase flow, (2) geysering, (3) oscillation due to hydrostatic head fluctuation, (4) density wave oscillation, (5) transition oscillation, and (6) stable two-phase flow. The main findings of the experiments are as follows: First, low amplitude geysering still occurs at 0.7 MPa under lower heat flux and high inlet subcooling. Second, stable two-phase natural circulation is achieved with system pressure as low as 0.2 MPa, under medium heat flux, and subcooling lower than 5 K. Third, oscillation due to hydrostatic head fluctuation only occurs under atmospheric condition. Finally, thermohydraulic stability maps and rational startup procedure are formulated.

A concept and safety characteristics of JAERI passive safety reactor (JPSR)

Abstract 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.

Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

Abstract The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept “RAPID-L” to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr. Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6Li as a liquid poison instead of B4C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented.

Critical heat flux experiments under steady-state and transient conditions and visualization of CHF phenomenon with neutron radiography

Abstract 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.

SCTF-III test plan and recent SCTF-III test results☆

Abstract A test plan of the Slab Core Test Facility with Core-III (SCTF-III) has been clarified. The previous SCTF-III tests simulating PWRs with combined-injection-type ECCS indicated the following results on the thermal-hydraulics in the full-radius core: 1. (1) two-region separation and multidimensional thermal-hydraulics both before and after reflood initiation. 2. (2) practically no core cooling in the region far from the water downflow region before reflood initiation in spite of good core cooling in the water downflow region. 3. (3) good core cooling even in the two-phase upflow region after bottom reflood initiation. 4. (4) large fall-back flow rate in comparison with the prediction by a typical one-dimensional correlation.

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 ...

Large-scale multi-dimensional phenomena found in CCTF and SCTF experiments

Abstract 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.

Utilization of rock-like oxide fuel in the phase-out scenario

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by nuclear material balance (NMB) analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the NMB code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels, and an ability to estimate waste and repository. Three scenario groups of once-through Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. Using two full-MOX or full-ROX reactors the Pu amount is reduced to about one-half and the isotopic vector of Pu deteriorated for being used as a nuclear weapon, especially in terms of spontaneous fission neutron generation. Effects of ROX reactors are more significant than MOX reactors in terms of both reduction in the Pu amount and deterioration of the isotopic vector. Repository footprint and potential radiotoxicity are not reduced by the MOX and ROX reactors because the heat and toxicity of MOX and ROX spent fuels are considerably high.

Transient Burnout under Rapid Flow Reduction Condition

Burnout characteristics were experimentally studied using uniformly heated tube and annular test sections under rapid flow reduction conditions. Observations indicated that the onset of burnout under a flow reduction transient is caused by the dryout of a liquid film on the heated surface. The decrease in burnout mass velocity at the channel inlet with increasing flow reduction rate is attributed to the fact that the vapor flow rate continues to increase and sustain the liquid film flow after the Inlet flow rate reaches the steady-state burnout flow rate. This is because the movement of the boding boundary cannot keep up with the rapid reduction of inlet flow rate. A burnout model for the local condition could be applied to the burnout phenomena with the flow reduction under pressures of 0.5∼3.9 MPa and flow reduction rates of 0.6∼35%/s. Based on this model, a method to predict the burnout time under a flow reduction condition was presented. The calculated burnout times agreed well with experimental resul...