Prototype Fast Breeder Reactor Monju Research Articles

Evaluation of the Shielding Design of the Primary Cooling System in the Prototype FBR Monju

As part of the start-up tests of the prototype fast breeder reactor Monju, the activities of 24Na and 22Na, which are generated by neutron irradiation of the primary sodium, were measured. The gamma-ray dose rates resulting from these nuclides inside the primary heat transport system cell were also measured. This paper presents an outline of the measurement methods and results, the design requirements and the design margins, which were evaluated by comparing the measurements with the calculations based on FBR shielding analysis methods.

Evaluation of Neutron Streaming and Future Shielding Measurement Plan in the Prototype FBR Monju

This paper describes the shielding measurements results in the reactor head access area (RHAA) and the primary heat transport system (PHTS) cells performed during the power raising test periods from February through December, 1995 in the prototype fast breeder reactor Monju. We confirmed that the measured values were well below the shielding design requirements in the RHAA and the PHTS cells and that the countermeasures against neutron streaming to the RHAA and the PHTS cells were very effective. Future measurements in order to obtain basic data for use in future FBR design are planned for the next period of operation.

「もんじゅ」温度計の流力振動

In December 1995, a thermocouple well was broken at the prototype fast breeder reactor Monju. In this paper, numerical simulation and water experiment of the flow-induced vibration of the thermocouple well are presented. Both the analysis and the experiment identified an in-line oscillation and the symmetric vortex shedding at the rated 100% flow condition. Furthermore, it has been found that the in-line oscillation takes place at decreased flow rate if the natural frequency of the thermocouple is reduced.

２６４． 高速増殖実証炉開発の現状と実用化見通し

The Demonstration Fast Breeder Reactor (DFBR) is the next step of FBR development following the prototype fast breeder reactor MONJU. The DFBR is now under development by The Japan Atomic Power Co. (JAPC) under the sponsorship of 9 Japanese electric power companies and Electric Power Development Co., Ltd. The JAPC has been performing the design study and R & D for DFBR in cooperation with Power Reactor and Nuclear Fuel Development Corp. (PNC), Central Research Institute of Electric Power Industry (CRIEPI) and Japan Atomic Energy Research Institute (JAERI). This report describes the prospect of FBR commercialization and the current status of new technology for DFBR and innovative technology toward FBR commercialization.

Irradiation performance of modified 316 stainless steel for Monju fuel

Neutron irradiation behavior of PNC316 stainless steel, which has been developed for prototype fast breeder reactor MONJU, was investigated up to a maximum dose of 185 dpa. Excellent swelling and creep properties of PNC316 were confirmed and are believed to be the best among this class of austenitic stainless steels. Performance of a Monju type fuel assembly was demonstrated at conditions more severe than those anticipated in Monju. Significant improvement in the swelling resistance of PNC316 was mainly due to the formation of a stable phosphide that traps helium bubbles and retards the conversion of bubbles to voids. Titanium or increased silicon addition apparently extended the dose for the phosphide dissolution and thus extended the swelling incubation period.

Simplified analysis and design for elevated temperature components of Monju

The Japanese Prototype Fast Breeder Reactor Monju is now at a stage of submitting applications to the government for getting Construction Permission. Detailed structural design of components is now in progress. As well understood, a number of components of Monju have to be subjected to elevated temperature design. For this need, Power Reactor and Nuclear Fuel Development Corp. has developed “Elevated Temperature Structural Design Guide for Class 1 Components of Prototype Fast Breeder Reactor” (ETSDG). It is now incorporated into “Construction Code for Prototype Breeder Reactor”. The basic concept of ETSDG is originated from ASME Boiler and Pressure Vessel Code Section III, Case Interpretation N-47, ETSDG prepares some extended simplified inelastic analysis methods in addition to the basic concept of N-47. This extension is necessary to make easy the complicated inelastic analyses. The simplified methods have been developed through many case studies with detailed inelastic analyses, elastic analyses and experimental verifications.