Prototype Fast Breeder Reactor Monju Research Articles

Evaluation of Reaction Rate Distribution for Shielding Region in the Prototype Fast Reactor Monju

The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The reaction rates of 238U (n,f) and 58Ni (n,p) sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values were slightly worse in the shielding regions; however, those for the 58Ni (n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the 238U (n,f) cross section in the energy range of around 700 eV. The reaction rates of 235U (n,f), 239Pu (n,f), 238U (n,γ), and 197Au (n,γ) sensitive to the lower-energy neutrons were mostly judged reliable. The data in the lower axial shielding region are less reliable but may be acceptable for the shielding calculation.

Evaluation of Reaction Rate Distribution Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, reaction rate distributions of the fission reaction rates of 239Pu, 235U, and 238U and the capture reaction rate of 238U were measured using activation foils during its system startup test. The measurements in the core and radial blanket regions were evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies were examined through a comparison with calculations. The reaction rate data measured in Monju were confirmed all reliable and useful as the validation data. The fission reactions of 239Pu, 235U, and 238U can be validated with an accuracy of a few percent in the core and blanket regions. The capture reaction of 238U in the core region also can be validated with a similar accuracy, whereas a precise calculation of the foil cross section is necessary to consider the resonance shielding effects of the surrounding fuel pins and a foil.

Estimation Method for Residual Sodium Amount on Unloaded Dummy Fuel Assembly

This study has researched an estimation method for the amounts of residual sodium film and sodium lumps on dummy fuel pins in the Japanese prototype fast breeder reactor Monju by fundamental experiments and demonstration experiments. The residual sodium amounts on the pin surface were measured using three types of test specimens: (a) single pin, (b) 7-pin assembly, and (c) 169-pin assembly. The single pin and 7-pin assembly experiments revealed that the withdrawal speed of the pins and improvement of the sodium wetting drastically increased the residual sodium amounts. Furthermore, the 169-pin assembly experiments measured the practical amounts of the residual sodium in the Monju dummy fuel assembly and demonstrated sodium draining behavior through small gaps between the pins. The estimation method includes four models such as a viscosity flow model, Landau-Levich-Derjaguin (LLD) model, an empirical equation related to the Bretherton model, and a capillary force model in a tube. These calculation results were comparable to the residual sodium amounts obtained by the experiments. In the tests of improving sodium wetting, the amounts of residual sodium on the test specimen were close to 1.4 times larger than those of the thin sodium film estimated by the LLD model. The increased amount of residual sodium by improving the sodium wetting was explained by the ratio of the adhesion energy ( γ S O − γ S L ).

Evaluation of Control Rod Interaction Measurement in the Prototype Fast Reactor Monju

For evaluating control rod (CR) interaction, CR worth values with and without the insertion of interaction CRs were measured in six CR interaction patterns by the continuous period method and the compensation method in the prototype fast breeder reactor Monju. The measured CR worth was evaluated in detail by conducting corrections, such as the thermal contraction of CR drive shafts, and compared with calculations to examine the reliability. The calculation-to-experiment values (C/Es) of examined CR worths with and without the interaction were 0.96 to 0.98 ± 0.03. The values of the measured CR interactions were obtained from those examined CR worth measurements as a change in target CR worth by the insertion of interaction CRs: 5.0 ± 0.6% to 10.1 ± 0.1%. The C/Es of the interaction in those cases were 0.78 to 0.93 within 5% to 27% uncertainty. Through this study, the consistency and reliability of the measurements and the calculations of the CR interaction were confirmed, and the measurements were judged useful.

Evaluation of Sodium Radioactivity in the Primary System of the Prototype Fast Reactor Monju

In this study, the sodium radioactivity of 24Na and 22Na in the primary system measured in the prototype fast breeder reactor Monju was evaluated, and the reliability of the measurements and calculations was examined. The calculated-to-experiment values and their uncertainties for 24Na and 22Na radioactivities were 0.97 to 1.07 and 8.1% to 11.0% and 1.03 to 1.16 and 23.3% to 24.1%, respectively, using the JENDL-4.0 nuclear data library. The 22Na radioactivity calculated with ENDF/B-VIII.0 was larger by 40% than those calculated with JENDL-4.0 and JEFF-3.3 due to the 23Na(n,2n) cross-section discrepancy. The importance of the 22Na neutron capture effect is also confirmed herein for the accurate evaluation of the 22Na radioactivity. The experimental data were judged to be useful for validating the calculation method for improving the reliability of future designs of sodium-cooled fast reactors.

Evaluation of Fuel Reactivity Worth Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, fuel reactivity worth was measured at six positions as the reactivity corresponding to the differences of critical control rod positions between cores with and without a dummy fuel subassembly. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data for fast reactor neutronics design methodologies are examined through a comparison with calculations. Calculated-to-experiment values and their uncertainty of fuel reactivity worth were 0.97 to 1.02 and 4% to 6%. Through this study, the consistency and reliability of the measurements and calculations were confirmed.

Evaluation of Fixed Absorber Reactivity Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, fixed absorber worth was measured as the difference of core reactivity measured by control rod worth between cores with and without a single fixed absorber or three fixed absorbers. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data are investigated through a comparison with calculations using the latest neutronics design methodology developed at Japan Atomic Energy Agency. Calculated-to-experiment values and their uncertainties of fixed absorber worth were 1.00 ± 0.05 and 1.02 ± 0.04, respectively. Through this study, the measurements and calculations were found consistent and reliable.

Numerical Investigation of Fluid – Structure Interaction (FSI) on Sodium Leakage Accident at Prototype Fast Breeder Reactor Monju

This study presents a numerical investigation of fluid-structure interaction about sodium leakage accident at prototype fast breeder reactor Monju. The dynamic forces of sodium liquid flow acting on the thermowell are the main cause of fatigue failure, which leads to the occurrence of thermowell cracking due to intense stresses and high fatigue-cycle. Since the location and magnitude of these stresses are unknown, an analysis of the vibrational characteristics and stresses caused by FSI was performed by using ANSYS to prevent similar accidents in the future. The most dangerous case was found for FSI analysis by comparisons between the natural frequency of thermowell and vortex shedding frequency at several operating conditions. The results showed that a stress concentration location is similar to the location of broken thermowell, and the amplitude of stress is large enough for fatigue damage.

A refined analysis on the power reactivity loss measurement in Monju

The present paper describes an evaluation of power reactivity loss data obtained from the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano et al, 2008) is updated in consideration of the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured values. The influences of these updates are quantitatively identified and the most precise and probable calculation-to-experiment (C/E) C/E value is derived and a thorough uncertainty evaluation is conducted. These reveal that analysis overestimates measurement by 4.6% with a measurement uncertainty of 2.0%. This discrepancy is reduced to as small as 1.2% when the core bowing effect is taken into account, which implies the importance of this effect in the calculation of power reactivity loss.

Measurement and analysis of feedback reactivity in the Monju restart core

The purpose of this study is to develop a feedback reactivity measurement technique in the Japanese prototype fast breeder reactor Monju and to validate calculation methodology to forecast the nuclear feedback phenomena. A feedback reactivity measurement technique has been developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (KR) and reactor vessel inlet temperature (Kin). This technique can precisely measure the two reactivity components simultaneously and was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties demonstrated that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The calculated and measured values of KR agreed within 1%, and the value of Kin was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2°C, which supports the validity of the temperature calculation.

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.

Performance evaluation of a resonance ionization mass spectrometer developed for the FFDL system of fast reactors

To prevent a fuel failure event from becoming a serious radiation accident, sodium-cooled fast reactors are equipped with a system for failed fuel detection and location (FFDL). The FFDL instrument employed in the prototype fast breeder reactor Monju is based on the gas tagging method, in which precise and accurate measurements of krypton and xenon isotope ratios (78Kr/80Kr, 82Kr/80Kr and 126Xe/129Xe) must be performed in a short time. Burnup measurements also contribute to accurate determination of 82Kr/80Kr. We have developed a highly sensitive resonance ionization mass spectrometer for the isotopic analyses, which uses resonance ionization of Kr and Xe atoms by a pulsed laser at wavelengths of 216.7 and 249.6 nm, respectively. In evaluating the performance of our spectrometer, we find that systematic errors caused by isotope shifts can be reduced to negligible levels, and that statistical errors of 3% at a nuclide concentration of 7 ppt can be achieved with a single measurement time of about 40 minutes for each Kr and Xe isotope ratio. This means that, within one hour, about 200 fuel assemblies can be individually identified with a probability of 99%, verifying the applicability of our spectrometer to the FFDL system of fast reactors.

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.

Criticality Evaluation for the Monju Restart Core

The Japanese prototype fast breeder reactor Monju restarted its system startup test in May 2010 after a 14-year interruption. In the first stage of the test, reactor physics parameters have been measured at a zero power level.The present paper describes the evaluation of the criticality data. The best-estimate value and its uncertainty are evaluated as accurately as possible, following the guidelines recommended by the International Criticality Safety Benchmark Evaluation Project.The restart core contains 1.5 wt% of 241Am, which is three times more than the previous test. To extract the influence of the 241Am accumulation on calculation accuracy, criticality data obtained in the previous test are evaluated at the same level of detail.The calculation accuracy is investigated with four major nuclear data libraries. It is confirmed that the accuracy is within 0.3% Δk/k, a 2σ value of experimental uncertainty, with JENDL-3.3, JENDL-4.0, and ENDF/B-VII.0. The reactivity change due to the 241Pu decay can be simulated within an accuracy of 1% with JENDL-4.0 and JEFF-3.1.

高速増殖原型炉もんじゅの新指針に照らした耐震安全性評価 その(4)２次主冷却系配管の耐震安全性評価

高速増殖原型炉もんじゅの新指針に照らした耐震安全性評価 その(4)２次主冷却系配管の耐震安全性評価

B116 高速増殖原型炉もんじゅ1次ナトリウム純化系プラギング計電磁ポンプ停止後再起動による配管合流部の熱過渡荷重による損傷評価(保全管理・補修/材料強度・破壊評価,OS6 保全・設備診断技術)

B116 高速増殖原型炉もんじゅ1次ナトリウム純化系プラギング計電磁ポンプ停止後再起動による配管合流部の熱過渡荷重による損傷評価(保全管理・補修/材料強度・破壊評価,OS6 保全・設備診断技術)

Nano-mesoscopic structure control in 9Cr–ODS ferritic steels

AbstractThe development status of 9Cr–oxide dispersion strengthened (ODS) ferritic steels was reviewed, focusing on the authors' activities. This material is the most prospective cladding of the advanced high burn-up fuel elements for Japanese prototype fast breeder reactor MONJU and international Generation IV advanced fast reactors, because of their inherent radiation resistance as well as superior high temperature strength. The Y2Ti2O7 complex oxide particles in nano-scale are precipitated through the dissolution of Y2O3 by mechanical alloying processing, followed by annealing at elevated temperatures >1273 K. The formation of the unique phase, residual alpha-ferrite, containing nano-oxide particles, can be controlled in terms of an amount of excess oxygen and titanium. The meso-scale residual alpha–ferrite formation induces the extremely highest creep rupture strength in this class of ferritic steel tubes. The new processing for cladding manufacturing of 9Cr–ODS ferrite steels was realised by using al...

Transport Criticality Analysis of FBR MONJU Initial Critical Core in Whole Core Simulation by NSHEX and GMVP

The Japanese prototype fast breeder reactor MONJU developmental project is of prime importance in the establishment of the commercialized fast breeder reactor technology in Japan. This work presents new results obtained by the 3D hexagonal transport nodal code NSHEX, which for the first time has been applied to MONJU criticality analysis in whole core simulation and up to 70 energy groups. Significant energy-group approximation effect on k eff caused by the collapsing into coarse energy-group structure has been found and then confirmed separately using the Monte Carlo analysis. On the basis of the presented applied results, a method of transport correction evaluation, based on 70 energy-group approximation can be recommended instead of the conventional few-group method that assumes inter-independency of energy-group approximation and transport effect.

Studies on Plutonium Burning in the Prototype Fast Breeder Reactor Monju

Studies have been performed on plutonium burning in the Japanese prototype fast breeder reactor Monju. The main aims of these studies were to illustrate the plutonium-burning capabilities of fast reactors and to investigate the consequences of the related core design measures on the main core characteristics of Monju. Burner cores with diluting pins, with diluting subassemblies (also called diluents), and with an internal slice of inert material have been investigated; these require an increased average plutonium enrichment and thus offer an enhanced plutonium-burning rate. On the other hand, the consequences of the elimination of the radial and/or axial blanket have been investigated.Among the burner concepts, the B4C containing diluents have been found to be preferable because they cause the smallest maximum linear rating increase and offer the largest flexibility to adapt their reactivity via a modification of the B4C content. They also do not require a new fuel subassembly concept.For the case of the blanket elimination, the replacement of the blankets by steel reflectors has been found to be the best solution. The main consequence of the elimination of both blankets is the increase of the maximum linear rating by up to 11%. Whether this increase may lead to problems will depend on the actual linear power level of the core.

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.