Prototype FBR Research Articles

Analysis of the Neutronic Properties of the Prototype FBR Monju Based on Several Evaluated Nuclear Data Libraries

In this work three major evaluated nuclear data libraries (JENDL-4.0, JEFF-3.1.2, ENDF/B-VII.1) are used to analyze the prototype FBR Monju. Cross section libraries were generated for the fast reactor analysis code ERANOS. The purpose is to validate our ERANOS libraries, the Monju core model and estimate the performance of the nuclear data sets. The result is that the libraries give adequate performance for applications where a reasonable accuracy is sufficient (e.g. conceptual design, scoping studies), with JENDL-4.0 giving the best results. The uncertainties of various reactor parameters are larger than the future target values for FBR analysis. For analysis with very high accuracy (e.g. detailed design, licensing), data adjustment is necessary.

S083013 もんじゅIVTM落下衝突挙動の動的解析([S083-01]原子力システムおよび要素技術)

Prototype FBR plant 'Monju' experienced IVTM (In-Vessel Transfer Machine) drop trouble on August 26, 2010. The collision damage and the impact load due to IVTM drop were investigated by the numerical dynamic response analysis. From the analysis results, it was confirmed that the impact load was significantly reduced by the plastic deformation of the installation flange of IVTM and the installation mat of the sleeve structure on the fixed plug, where these two parts were directly contacted each other at collision, compared with the results by the hypothetical elastic analysis. And also it was confirmed that the plastic regions were limited to near collision surface of these two parts and almost other parts remained elastic. This means that the two parts which are contacted each other at collision should be replaced, but other parts are little damaged and are expected to be reused.

Uncertainty analysis of the prototype FBR Monju with the JENDL-4.0 nuclear data set

This paper discusses the uncertainty analysis of the criticality of the prototype Fast Breeder Reactor Monju with JENDL-4.0 nuclear data. The JENDL-4.0 nuclear data are more physically correct and complete than previous JENDL releases. The well-established ERANOS code was used. Since no cross section libraries based on JENDL-4.0 exist for ERANOS, these libraries were generated as part of the work. To validate the newly made libraries, benchmark calculations were performed. A model of the Monju reactor was prepared, based exclusively on publicly available information. The criticality of May 2010 of Monju was adequately calculated. A perturbation analysis identified the most important differences between JENDL-3.3 and JENDL-4.0. A strong contribution of 241Am was found. In the subsequent uncertainty analysis, it was found that the overall uncertainty is roughly the same between JENDL-3.3 and JENDL-4.0. This result was not as expected. Two main causes have been identified: firstly, in JENDL-4.0 covariance data has been added for isotopes and energy ranges for which there previously was no covariance data. Secondly, for some isotopes the covariances in JENDL-4.0 are larger than in JENDL-3.3. For several main reactions and isotopes, the uncertainty is largely reduced between JENDL-3.3 and JENDL-4.0. But for many isotopes with a non-negligible contribution the uncertainty is increased in JENDL-4.0. The balance is a slight increase in the overall uncertainty when JENDL-4.0 is used.

Application of the modified neutron source multiplication method to the prototype FBR Monju

The Modified Neutron Source Method (MNSM) is applied to the prototype FBR Monju. This static method to estimate the reactivity in sub-critical conditions has already given good results on commercial pressurized water reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement between different reactor states. Among Monju’s particularities that have a big influence on the MNSM factors are: the presence of two californium sources near the core and the position of the detector, which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and (α, n) reactions, which have increased during the shutdown period of 15years, will also be discussed. In order to evaluate the detector count rate, a propagation calculation has been conducted from the reactor vessel to the external detector. For two sub-critical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). Results indicate a good agreement between the MNSM reactivity and the reactivity measured with other methods. The reactivity dependence of the correction to apply to the point kinetic equation is discussed.

Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor

This paper discusses the objectives and results of a multiyear R&D project to improve the modeling accuracy for the detailed calculation of the Japanese Sodium-cooled Fast Reactor (JSFR), although the preliminary design of JSFR is prepared using conventional methods. For detailed design calculations, new methods are required because the JSFR has special features, which cannot be accurately modeled with existing codes. An example is the presence of an inner duct in the fuel assemblies. Therefore, we have developed new calculational and experimental methods in three areas: (1) for neutronics, we discuss the development of methods and codes to model advanced FBR fuel subassemblies, (2) for fuel materials, modeling and measurement of the thermal conductivity of annular fuel is discussed, and (3) for thermal hydraulics, we describe advances in modeling and calculational models for the intermediate heat exchanger and the calculational treatment of thermal stratification in the hot plenum of an FBR under low flow conditions. The new methods are discussed and the verification tests are described. In the validation test, measured data from the prototype FBR Monju is partly used.

Resumption of System Start−up Test of Prototype FBR Monju

Resumption of System Start−up Test of Prototype FBR Monju

RELAP5/3Dによる「もんじゅ」蒸気発生器タービントリップ試験解析

RELAP5/3Dによる「もんじゅ」蒸気発生器タービントリップ試験解析

３次元輸送計算コードTORTの「もんじゅ」しゃへい解析への適用性検討(その５)-１次主冷却系配管しゃへい室-

３次元輸送計算コードTORTの「もんじゅ」しゃへい解析への適用性検討(その５)-１次主冷却系配管しゃへい室-

高速増殖原型炉「もんじゅ」２次系床ライナの健全性評価

高速増殖原型炉「もんじゅ」２次系床ライナの健全性評価

Controversial Technical Issues on Nuclear Safety of Prototype FBR Monju Contained in the Decision by the High Court

Controversial Technical Issues on Nuclear Safety of Prototype FBR Monju Contained in the Decision by the High Court

1142. 異常高温時の内圧負荷を受ける伝熱管の破裂挙動

Rapid bursting tests using the Tube Rupture Simulation Test Rig (TRUST-2) and its overheating rupture analyses were carried out on the 2·1/4Cr-1Mo steel tube under an internal pressure load with extraordinarily high temperature condition. The tube failure behaviors were classified into three patterns of the ductile failure, the creep failure, and the ductile failure accompanied by creep, depending on the test conditions. In comparison between the test and the analysis on the creep failure and the ductile failure accompanied by creep, the time to failure of the analysis is estimated 25-50% shorter than that of the test, and the analysis is considered to be conservative. On the ductile failure, the times to failure of the test and the analysis become close but that of the analysis is still estimated about 10% shorter; the temperature to failure of the analysis is calculated about 100°C lower than that of the test; therefore, the conservatism of the analysis is confirmed. On the creep failure, if the time coefficient of the analysis is given about 1.5-2, the time to failure of the analysis becomes close to that of the test, which means that the analysis by use of the time coefficient 3 results in having the safety margin of 1.5-2 in the application to the Prototype FBR MONJU.

Measurement of Shielding Characteristics in the Prototype FBR Monju

In the prototype fast breeder reactor Monju, shielding measurements were made around the reactor core, the primary heat transport system (PHTS), and the fuel handling and storage system during the system start-up tests at different power levels between 0% and 45%. The objectives of the tests were to evaluate the margins by which the shielding performance exceeds the original design requirements, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. This paper summarizes the important features of the Monju shielding structures and the shielding measurements.

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 the Shielding Design around the Reactor Core in the Prototype FBR Monju

This paper describes shielding evaluation of the measurements around the reactor core of Monju. The measurements were performed during the system start-up tests at different power levels between 0% and 45%. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis methods. Based upon these results, the design margins around the reactor core have been re-examined and re-confirmed.

Evaluation of the Shielding Design of the Fuel Handling and Storage System in the Prototype FBR Monju

Shielding performance tests of the fuel handling and storage system were carried out at Monju in October 1993. The performance of the shielding of the fresh fuel storage rack against neutron and gamma radiation from fresh fuel assemblies, that of the invessel fuel transfer machine (IVTM) and the ex-vessel fuel transfer machine (EVTM) against neutron and secondary gamma-rays generated in the shielding or structural materials from the neutron source assembly (252Cf), were assessed. Dose rates were measured using rem-counters and solid-state track-detectors (SSTDs) for neutrons and ionization chamber survey-meters and film dosimeters for gamma-rays. 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.

The present status of DFBR development in Japan

The first demonstration fast breeder reactor (DFBR-1) is being developed in Japan, as the essential step toward the commercial deployment of FBR around 2030. The construction of the DFBR-1, which follows the experimental FBR (JOYO) and the prototype FBR (MONJU), is scheduled for the early 2000s. The Japan Atomic Power Company (JAPC) has been conducting design studies of the DFBR-1 and relevant R&D since 1986, under a contract awarded by the nine electric companies and the Electric Power Development Co., Ltd. The electric utilities have settled the major design specifications of the DFBR-1 plant as the top-entry loop type system with an output of 660MWe. Based on these major design specifications, JAPC has been conducting the optimization design study of the DFBR-1 since FY 1994 in order to establish the overall consistent plant design with construction cost reduction below 1.5 times that of LWRs on 1,000 MWe basis and comfirm the licensability of the DFBR-1 design. This study will complete by the end of FY 1996 This paper mainly outlines the design concepts of DFBR-1 resulting from the optimization design study of DFBR-1.

Transport of Fresh Mox Fuel Assemblies for the Monju Initial Core

AbstractTransport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of, the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occur...

Transport of Fresh MOX Fuel Assemblies for the MONJU Initial Core

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of, the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.

Development of the demonstration FBR and its commercialization

The FBR being developed in Japan is expected to operate commercially by the year 2030 or so. The construction of the first demonstration FBR “DFBR-1”, which follows the experimental FBR “JOYO” and the prototype FBR “MONJU” is scheduled for early 2000s. The Atomic Energy Commission adopted these policies in the new version of “The Long Term Program for Research, Development and Utilization of Nuclear Energy”. The Japan Atomic Power Company has progressed design studies of the DFBR-1 and R&Ds. Electric utilities settled the major design specifications of the DFBR-1 such as the top entry loop type as the reactor type and about 660 MWe of plant capacity. To achieve the goals for the commercialization of FBRs, such as a high level of safety and reliability comparable with a LWR and further improvement of economy, many innovative technologies and system improvements will be required in commercial FBRs. This paper describes the present status of development of the DFBR-1.