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Journal Articles

EBR-II passive safety demonstration tests benchmark analyses; Phase 2

Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08

The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.

Journal Articles

Inter-subassembly heat transfer of sodium cooled fast reactors; Validation of the NETFLOW code

Mochizuki, Hiroyasu

Nuclear Engineering and Design, 237(19), p.2040 - 2053, 2007/10

 Times Cited Count:21 Percentile:16.59(Nuclear Science & Technology)

This paper describes an applicability of the NETFLOW code for prediction of temperature at the exit of subassemblies of a sodium cooled fast reactor. So far, this code has been validated using data obtained at facilities and reactors of which coolants are water or sodium. A natural circulation test was conducted in the "Joyo" reactor with a 100 MW irradiation core, and a turbine trip test at "Monju". These tests were chosen to validate a model to calculate inter-subassembly heat transfer. Through a calculation for the natural circulation of "Joyo", a model to calculate heat transfer in radial direction of the inter-subassemblies gave the reasonable sodium temperatures at the exit of subassemblies. Good agreements were also obtained in prediction of temperatures of "Monju". It was shown that the NETFLOW could evaluate temperatures at the exit of the subassemblies with the inter-subassembly heat transfer model without heat transfer by inter-wrapper flow.

Journal Articles

Development of a versatile plant simulation code with PC

Mochizuki, Hiroyasu

Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 9 Pages, 2007/05

This paper describes about the development of the plant simulation code NETFLOW. The code developed to simulate thermal-hydraulics in the auxiliary cooling system of a reactor has expanded to simulate plant dynamics not only for light water reactors but also for liquid metal cooled fast reactors (LMFRs). In order to improve the applicability for LMFRs, models affecting on characteristics of natural circulation are improved. They are a model of inter-subassembly heat transfer and a model of heat transfer coefficient of finned heater tubes in an air cooler. A natural circulation in the Joyo experimental fast reactor after the reactor trip is calculated, and good agreement has been obtained between the test and simulation results.

Journal Articles

Analysis of the Chernobyl accident from 1:19:00 to the first power excursion

Mochizuki, Hiroyasu

Nuclear Engineering and Design, 237(3), p.300 - 307, 2007/02

 Times Cited Count:6 Percentile:53.42(Nuclear Science & Technology)

Many researchers have reported that the root cause of the Chernobyl accident has not been clarified still now. Since most of them discussed the accident without a precise thermal-hydraulic investigation, thermal-hydraulic calculations coupled with neutronic calculations have been done on the basis of the recorded result at the Chernobyl Unit 4. Calculation could trace plant parameters from 1:19:00 to the first power excursion. Reactivity slightly smaller than 1$$beta$$ by the positive scram is a possible direct cause of the accident, which acts as a trigger to increase the reactor power.

Journal Articles

Verification of NETFLOW code using plant data of sodium cooled reactor and facility

Mochizuki, Hiroyasu

Nuclear Engineering and Design, 237(1), p.87 - 93, 2007/01

 Times Cited Count:9 Percentile:40.6(Nuclear Science & Technology)

The plant simulation code NETFLOW on PC applicable to the liquid-metal cooled reactors has been developed on the basis of the models developed for single-phase and two-phase light water flow systems. The functions of this code have been verified by individual tests for light water flow systems and a sodium flow system. In order to apply this code to a sodium cooled fast reactor, several extra functions were verified using the plant data obtained using 50 MW steam generators and the Monju fast breeder reactor. Finally the turbine trip transient of the Monju was simulated and the result was compared with the measured plant data. Good agreements were obtained in these verifications. As a result of the present study, the code can be applied as an education tool for students.

JAEA Reports

Plant Dynamics Analysis using NETFLOW Code

Mochizuki, Hiroyasu

JNC-TN9400 2005-004, 44 Pages, 2005/04

JNC-TN9400-2005-004.pdf:1.3MB

The NETFLOW code that can be used for liquid metal flow system has been developed on the basis of the code that can be used for a single-phase and two-phase light-water flow systems realized in pressure-tube-type reactors and LWRs. The algorithm of the code for the liquid-metal flow is developed referring to the Super System Code (SSC) in order to develop further a new code that can solve stably the thermal hydraulics with high speed. Since the electric manual of the code can be referred on a PC, one can make an input data and run on the same PC.The functions of this code were verified by individual tests for light water systems. There are a lot of experiences in applications for power plants and thermal-hydraulic experimental loops. The present report evaluates the applicability of the code and clarifies problems. Experimental examples for the evaluation are steady state test using 50 MW Steam Generator, natural circulations experiments using the PLANDTL facility, a natural circulation experiment and a turbine trip experiment of the Monju reactor.The code could generally trace these experimental results, and simulate the plant transient thermal-hydraulic behaviors of liquid metal coolant approximately 1000 times faster than the real phenomena. However, further modification should be necessary in order to improve the prediction accuracy.

JAEA Reports

THE O-ARAI FR CYCLE SYMPOSIUM 2004; The direction of the world in FR cycle development and the role of "Joyo" and "Monju" reactors

Koi, Mamoru; Mochizuki, Hiroyasu

JNC-TN9200 2004-001, 335 Pages, 2004/06

JNC-TN9200-2004-001.pdf:35.21MB

The O-arai Fast Reactor (FR) cycle symposium 2004 was held on the theme of the direction of the world in FR cycle development and the role of "Joyo" and "Monju" reactors in the O-arai Engineering Center (OEC) on 27 February 2004. Approximately 400 people including municipal people, specialists of FR cycle technology development in Japan and other countries, students, etc attended the symposium.

JAEA Reports

Large-scaled thermohydraulic tests plan for cooling systems in fast reactors; Experimental models of reactor vessel and the primary cooling system

Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.

PNC-TN9410 96-279, 51 Pages, 1996/08

PNC-TN9410-96-279.pdf:2.92MB

Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.

Journal Articles

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

Mochizuki, Hiroyasu; Koike, Mitsutaka;

Nuclear Engineering and Design, 144(2), p.293 - 303, 1993/10

None

Journal Articles

Experimental and Analytical Studies of Flow Instabilities in Pressure Tube Type Heavy Water Reactors

Mochizuki, Hiroyasu

Journal of Nuclear Science and Technology, 29(1), 50 Pages, 1991/12

None

Journal Articles

Development of ATR Type Steam Separator Performance Analysis Code and its Validation, 1; Carryunder Characteristic

Mochizuki, Hiroyasu; Hirao, Yasuhiko

Journal of Nuclear Science and Technology, 28(12), p.1078 - 1089, 1991/00

None

Oral presentation

Study on heat transfer of air cooler for fast reactors

Takano, Masahito*; Mochizuki, Hiroyasu

no journal, , 

An empirical heat transfer correlation of air on finned heater tubes was derived on the basis of experimental results obtained in air coolers (outside: air, inside: sodium) of fast reactor. Real capacity of the air cooler at Monju was evaluated by taking into account of the fin efficiency into the correlation. A transient analysis was aloso conducted using a plant dynamics analysis code.

Oral presentation

Study on heat transfer in intermediate heat exchangers and natural circulation of fast reactors

Takano, Masahito*; Mochizuki, Hiroyasu

no journal, , 

An intermediate heat exchanger (IHX) of a sodium cooled fast reactor is a component to transfer heat generated in the core to the secondary heat transport system. In the present study, an empirical correlation about heat transfer coefficient has been discussed using test data of IHXs at the Joyo with the MK-III core, at the 50MW-SG facility, and also at the prototype reactor Monju in order to apply it to plant analysis of Monju, and the system and component design of the future reactor. After the obtained correlation is incorporated into the plant dynamics analysis codes, it has been confirmed that a test of the natural circulation in the secondary heat transport system of Monju can be tranced by the codes.

13 (Records 1-13 displayed on this page)
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