Yoshihisa Nishi

Current Design Status of Sodium Cooled Super-Safe, Small and Simple Reactor

CRIEPI has been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. And a conceptual design of 4S (Super-Safe, Small and Simple) reactor was proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void coefficient are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core life time is more than 10 years; (5) Its construction, maintenance and operation are expected to be very simple by eliminating active components from inside of a reactor vessel. The 4S reactor is a sodium cooled fast reactor and its reactivity is not controlled by neutron absorber rods but by neutron reflectors. An electrical output is 50 MW. This paper describes a design modification to enhance the feasibility from the previous 4S design. A core active height can be shortened to 1.5 m from 4.0m to keep the reactivity characteristics. An averaged fuel burn-up is up to 70 GWD/ton and a pressure drop at the core region is less than 0.1 MPa. A reactivity control system is modified according with the core design change. As for the steam generator design, sodium-water reaction accidents must be taken into account as a design basis event for the utilization of the secondary sodium coolant. Therefore, a modified plate type heat exchanger is proposed as a steam generator. It may be possible to develop a compact steam generator, which is free from sodium-water reaction accidents and to eliminate the secondary sodium systems. The 4S reactor without secondary system has been proposed as a candidate design.Copyright

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (IV) - Assessment of hydrogen behavior in stored Cs adsorption vessel

Hydrogen diffusion behavior in a cesium adsorption vessel is assessed. The vessel is used to remove radioactive substance from contaminated water, which is proceeded from Fukushima accident. Experiment and numerical calculation are conducted to clarify the characteristics of natural circulation in the vessel. The natural circulation arising from the temperature difference between inside and outside the vessel is confirmed. We develop an evaluation model to predict the natural circulation and its prediction agrees well with the results obtained by the experiment and the calculation. Using the model, we predict steady and transient behavior of hydrogen concentration. Results indicate that hydrogen concentration is kept lower than the flammability limit when the short vent pipe is open.

Verification of the Plant Dynamics Analytical Code CERES Using the Results of the Plant Trip Test of the Prototype Fast Breeder Reactor MONJU

CERES is plant system analysis code for LMRs (liquid metal cooled reactors) developed by the Central Research Institute of Electric Power Industry (CRIEPI). To verify the CERES code, analyses were performed by using the result of the plant trip test of the prototype FBR (fast breeder reactor) “MONJU” at 40% rated power. The verification work was performed as a joint research of CRIEPI and JAEA (Japan Atomic Energy Agency). Following three verification analyses were performed mainly. (I) Analysis concerning the primary/ secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) was modeled in R-Z 2-dimension). (II) Analysis concerning the thermal-hydraulic characteristics in the plenum of R/V (the plenum was modeled in 3-dimension). (III) Analysis concerning the flow characteristics inside the intermediate heat exchanger (IHX) (the plenum in the IHX was modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the “MONJU”. Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plena of “MONJU” became clear by these analyses.Copyright

Passive Safety Features in Sodium Cooled Super-Safe, Small and Simple Reactor

This paper describes the passive safety features utilized in the updated sodium cooled Super-Safe, Small and Simple fast reactor, which is the improved 4S reactor. This reactor can operate up to ten years without refueling and neutron reflector regulates the reactivity. One of the design requirements is to secure the core against all anticipated transients without reactor scram. Therefore, the reactor concept is to design to enhance the passive safety features. All temperature reactivity feedback coefficients including whole core sodium void worth are negative. Also, introducing of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. Safety analyses are carried out to simulate various transient sequences, which are loss of flow events, transient overpower events and loss of heat sink events, in order to evaluate the passive safety capabilities. A calculation tool for plant dynamics analyses for fast reactors has been modified to model the 4S including the unique plant system, which are reflector control system, circulation pumps and RVACS. The analytical results predict that the designed passive features improve the safety in which temperature variation in transients are satisfied with the safety criteria for the fuel element and the structure of the primary coolant boundary.Copyright

Effect of Two-Phase Flow Structure in Decontamination Factor of Filtered Containment Venting System

In order to gain the best use of filtered containment venting systems (FCVSs), the decomtamination factor of FCVSs is to be investigated as a function of system parameter including steam flow rate, pressure, temperature, water level, and operating time. A full-height test facilities were designed and constructed in Central Research Institute of Electric Power Industry (CRIEPI), Japan to evaluate the decontamination factor (DF) in FCVSs. The target types are the orifice and the venturi FCVSs. The height and the internal diameter of the cylindrical test vessel is 8 m and 0.5 m. Bubbly flows were visualized through the view window up to 0.8 MPa and 170 °C. Steam bubbles in 0.2 wt% sodium thiosulfate and 0.5 wt% sodium hydroxide were found to be much smaller than those in water. The DF were evaluated for the aerosol, elemental iodine and organic iodine. The installed aerosol optical spectrometer measures the number density and the diameter of aerosols. The concentrations of elemental iodine were quantified with an inductively-coupled plasma with mass spectrometry (ICP-MS). The concentration of organic iodine was quantified with a gas chromatography with mass spectrometry (GC-MS). In order to investigate two-phase flow dynamics in the vessel, separate effect tests were conducted with air-water test facility. The height of cylindrical test vessel is 8 m. Visual observation was conducted for two internal diameter levels: 0.05 and 0.5 m. High speed video frames were recorded through the transparent (acrylic) vessel wall. Wire-Mesh Sensors (WMS) were installed to acquire a cross-sectional void fraction to compare with DF in the facility. On the basis of the obtained database, we develop the FCVSs performance evaluation technique and propose an optimal FCVSs operation method for a further safety improvements of the nuclear power plant.© 2014 ASME

Computational analysis of the thermal-hydraulic characteristics of the encapsulated nuclear heat source

Abstract The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 U.S. Department of Energy Nuclear Energy Research Initiative program as a candidate Generation IV reactor concept. It is a fast neutron spectrum reactor cooled by lead-bismuth eutectic using natural circulation. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant through rectangular intermediate heat exchanger (IHX) channels. The decay heat is removed by the reactor vessel auxiliary cooling system (RVACS). Events of protected loss of heat sink (PLOHS) and unprotected transient overpower (UTOP) have been analyzed for the ENHS using the CERES transient simulation code for liquid-metal-cooled reactors. It is found that the ENHS core is sufficiently cooled by the RVACS under the PLOHS condition. The core flow rate is affected by the growth and disappearance of temperature stratification in the primary plenum. It is also found that even under the inconceivable UTOP event considered, the ENHS reactor core is not catastrophically damaged. This is due to negative reactivity feedback from the radial expansion of the core, the grid plate, and the Doppler effect. The use of high-performance ferritic steel instead of HT-9 and proper design of the reactor control system could provide large safety margins against cladding damage.

A New Concept of the 4S Reactor and Thermal Hydraulic Characteristics

CRIEPI (Central Research Institute of Electric Power Industry) has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application to dispersed energy supply and multipurpose use, with Toshiba Corporation [1,2,3,4]. Electrical output of the 4S reactor is from 10MW to 50MW, and burn-up reactivity loss is regulated by neutron reflectors. The reflector that surrounds the core is gradually lifted up to control the reactivity according to core burn-up. 30year core lifetime without refueling can be achieved with the 10MW 4S (4S-10M) reactor. All temperature feedback reactivity coefficients, including coolant void reactivity, of the 4S-10M are negative during the 30year lifetime. A neutron absorption rod is set at the center of the reactor core with the ultimate shutdown rod. The neutron absorption rod used during the former 14 years is moved to the upper part of the reactor core, and the operation is continued through the latter 16 years. The pressure loss of the reactor core is lower than 2kg/cm2 to enable effective utilization of the natural circulation force, and the average burn-up rate is 76GWD/t. To suppress the influence of the scale disadvantage, loop-type reactor design is one of the candidates for the 4S-10M. The size of the reactor vessel can be miniaturized by adopting the loop type design (4S-10ML). In the 4S-10ML design, integrated equipment which includes primary and secondary electromagnetic pumps (EMPs), an intermediate heat exchanger (IHX) and a steam generator (SG) is adopted and collocated by the reactor vessel. The decay heat removal systems of 4S-10ML consist of the reactor vessel air cooling system (RVACS) and SGACS (a similar system to the RVACS, with air cooling of the outside of the integrated equipment vessel). They are completely passive systems. To decrease the construction cost of the reactor building, a step mat structure and the horizontal aseismic structure are adopted. 4S-10ML has unique features in the cooling systems such as integrated equipment and two separate passive decay heat removal systems which operate at the same time. To evaluate the design feasibility, the transition analyses were executed by the CERES code developed by CRIEPI [5]. In this paper, the design concept of 4S-10ML, and the results of the plant transition analyses are described.© 2004 ASME

An evaluation model to predict steam concentration in a BWR reactor building

When there is no power for cooling the spent fuel pool and conditioning the air in a boiling water reactor (BWR) reactor building, water vapor is generated from the pool and it affects the atmosphere in the building. To consider the impact of the steam in preparing emergency operation procedures, the building atmosphere under various conditions is to be evaluated with reasonably low computational cost. A lumped parameter model to predict the transient behavior of the building atmosphere was developed, in which the evaporation from the spent fuel pool and the condensation to the wall were taken into consideration. A transient behavior of temperature and vapor concentration in a BWR operating floor was predicted with the model. The results and the prediction speed were compared to those of a three-dimensional computational fluid dynamic calculation, and it was confirmed that the model could obtain almost the same results about 280,000 times faster. Parameter studies are conducted with the model, and dominant parameters to the evaporation and the condensation were clarified.

Development of a Multi-Dimensional Measurement Sensor of Void Fraction and Phasic Velocity for Boiling Two-Phase Flow in a 5×5 Heated Rod Bundle

A subchannel void sensor (SCVS) was developed to measure the cross-sectional distribution of a void fraction in a 5×5 heated rod bundle with o.d. 10 mm and heated length 2000 mm, and applied in a boiling two-phase flow experiment under the atmospheric conditions assumed in an accident and spent fuel pool. The SCVS comprises 6-wire by 6-wire and 5-rod by 5-rod electrodes. Wire electrodes 0.2 mm in diameter are arranged in latticed patterns between the rod bundle, while a conductance value in a region near one wire and another gives a local void fraction in the central-subchannel region. 32 points (= 6×6−4) of the local void fraction can be obtained as a cross-sectional distribution. In addition, a local void fraction near the rod surface can be estimated by a conductance value in a region near one wire and one rod using the simulated fuel rods as rod electrodes, which allows 100 additional points (=4×25) of the local void fraction to be acquired. The devised sensors are installed at five height levels to acquire two-phase flow dynamics in an axial direction. A pair of SCVS is mounted at each level and placed 30 mm apart to estimate the one-dimensional phasic velocity distribution based on the cross-correlation analysis of both layers. The time resolution of void measurement exceeds 800 frames (cross-sections) per second. The heated rod bundle has an axially and radially uniform power profile, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The boiling two-phase flow experiment, which simulated a boil-off process, was conducted with the devised SCVS and experimental data was acquired under various experimental conditions, such as inlet-flow velocity, rod-bundle power and inlet subcooling. The experimental results exhibited axial and radial distribution of two-phase flow structures, i.e. void-fraction and phasic-velocity distributions quantitatively.Copyright

Validation of TRACE Code for Flashing-Induced Density Wave Oscillations in SIRIUS-N Facility, Which Simulates ESBWR

The TRACE code was validated against the flashing-induced density wave oscillation in the SIRIUS-N facility at low pressure (from 0.1 to 0.5 MPa) as a part of the international CAMP-Program of USNRC. The SIRIUS-N facility is a scaled copy of natural circulation BWR (ESBWR). Stability map of TRACE agrees with that of SIRIUS-N facility at low subcooling region, though instability observed in the lower heat flux and higher subcooling region from the stability limit of experiment. The TRACE code demonstrates the flashing-induced density wave oscillation characteristics: The oscillation period correlates well with the transit time of single-phase liquid in the chimney regardless of the system pressure, inlet subcooling, and heat flux. Unlike Type-I and II density wave oscillations, the inlet or exit throttling does not affect stability boundary and oscillation amplitude of flashing-induced density wave oscillations significantly. Increasing pressure decreases oscillation amplitude. The comprehensive validation confirms that the TRACE code can demonstrate thermal-hydraulic stability of natural circulation BWRs.Copyright