Experimental Fast Reactor JOYO Research Articles

Model-Based Temperature Noise Monitoring Methods for LMFBR Core Anomaly Detection.

Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the model has a higher precision than the physical model in background noise prediction. Based on these results, an autoregressive model modification method is proposed, in which a temporary model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio.

Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR

Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOYO was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation (EPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-in-duced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425°C to a neutron fluence of 8.3 × 10 26 n/ m 2 (〉 0.1 MeV) which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300°C, but was not observed on specimen tested in water at 60°C and in air at 300°C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCO in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum.

Characterization of microstructure at 673 K in neutron-irradiated Fe-16Ni-15Cr with the addition of C, Ti and Nb

Simple model alloys of Fe-16 Ni-15 Cr- xC- yTi- zNb were irradiated at 673 K up to 2.5 dpa in the experimental fast breeder reactor JOYO. TEM examinations were made to investigate the separate and synergistic effects of C, Ti and Nb additions on the development of microstructure. Additions of C decreased swelling resulting from decreased cavity number densities. Ti or Nb alone changes little swelling. Ti significantly decreased cavity number densities. C additions to the alloys with Ti and Nb increased the size and number density of cavities; the number density of smaller sized cavities (for example, < 5 nm in diameter) was remarkably increased. Faulted dislocation loops were characteristically formed by additions of Ti and Nb, and became smaller in size and much more circular in shape with fluence. This tendency was enhanced by the further addition of C, where unfaulted loop or dislocation tangling was rarely observed and no precipitation of TiC or NbC occurred.

Experiment and analysis on natural convection characteristics in the experimental fast reactor joyo

A series of decay heat removal tests by natural convection were carried out at various decay heat levels in the experimental fast reactor JOYO to study thermohydraulic behavior in fast breeder reactors (FBR). The principal results obtained are as follows: 1. (1) It was proved that the decay heat could be safely removed by natural convection even at 100 % full power (100 MWt). 2. (2) A plant-wide dynamics code, MIMIR-N2, was developed and verified through experimental data in order to establish an analysis technique for evaluating natural convection behavior. 3. (3) Verification data for assessing the natural convection ability in future plants were obtained through this test.

Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging

Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1 × 1024 n/m2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M6C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M6C in JPCA-2.

Development of quality control data base “DANTE” and relational analysis between fission gas release behavior and plutonium homogeneity in mixed oxide fuel by DANTE

The PNC Plutonium Fabrication Facility (PPFF) of Tokai Works of the “Power Reactor and Nuclear Fuel Development Corporation” (PNC) has produced several types of mixed oxide fuels which have been used as the fuels for the Advanced Thermal Reactor FUGEN, and the Experimental Fast Reactor JOYO. Accumulated production of MOX fuel is about 58000 pins as of December, 1986. PPFF production technique is so reliable that no failure of driver fuel pins has been found in the FUGEN and JOYO. This is based on the adequate process control and quality control. In this paper the outline of quality control method applied to MOX fuels in our facility will be reviewed. The discussion is done mainly on the viewpoint of the software system including data gathering, data analysis and data storage, etc. The code system, Data Analysing Technology for Engineers, DANTE, has been developed for this purpose. The DANTE code system is consisted of real time system having relational data base. It is possible to refer the data base relationally and to make tables and figures of the analyzed results on real time by using the DANTE code system. The DANTE code system contributes to quality assurance in the MOX Fuel production facility by evaluating overall data which come from each process such as production and inspection processes.

しゃへい型イオンマイクロアナライザ（ＳＩＭＡ）による燃焼度測定技術の開発

A new method of the fuel burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. This method is based on the direct isotope analysis of U, Pu and Nd in an irradiated fuel by means of the secondary ion mass spectrometry. The sample tests on the irradiated fuels up to 13a/0 burnup have shown an accuracy within 6%.While the conventional isotope dilution method requires the acid-dissolution and the chemical separation of the irradiated fuel by handling samples in a glovebox, the new method can measure the burnup on a fuel in shielded apparatus remotely and directly. Consequently, the time for measurement was reduced to about one tenth of that of the conventional method, and a human radiation exposure became negligible.A microarea surface analysis by the SIMA provides the local burnup distribution in small area. Using this technique, the radial profile of the burnup distribution due to a thermal neutron depression was obtained in the fuel pellet irradiated in JRR-2 reactor. The radial burnup distribution was also measured in the mixed-oxide fuel pellet of the experimental fast reactor JOYO MK-II, which revealed some effects of Pu redistribution on the radial burnup profile.

Development of a New Measurement Method for Fast Breeder Reactor Fuel Burnup Using a Shielded lon Microprobe Analyzer

A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis.

Waterhammer in a Primary Coolant System by Controlling a Circulating Pump Revolution

Studies have been made to compare, and discuss, the analytical solutions obtained by the method of characteristics with the experimental results about the waterhammer. The measurements have been carried out using a 1/6 scale model of the primary coolant system of the experimental fast breeder reactor JOYO belonging to the Power Reactor and Nuclear Fuel Development Corporation. The waterhammer has taken place at the valve throat area ratio ranging from 1/4 to 4/4 with variations in circulating pump revolving speed in the angular acceleration rages 20 to 180 rad/s2 and -20 to -180 rad/s2. Good agreement is found between the analytical and experimental results. Therefore the calculation models and idea used in this paper are practically useful enough.

高速炉照射による316ステンレス鋼のボイドスエリングと組織変化

Void swelling and microstructural change during irradiation were investigated for Type 316 stainless steels different in the fabrication process and/or the phosphorus content. The relation between Ni content and swelling was also examined. The irradiations were performed in the Japan Experimental Fast Reactor JOYO, the French fast reactor Phenix, etc.The irradiated specimens were examined by a scanning transmission electron microscope equipped with an energy dispersive X-ray analyser. Results obtaind are summarized as follows.(1) Phosphorus in the steel was found to be effective as a swelling inhibitor.(2) Cold work reduced the swelling of the steel.(3) The segregation of Ni to grainboundaries, precipitates and surroundings of voids decreased the Ni content in the austenite matrix.(4) With increaseing swelling, the initial Ni content of 13% decreased to approximately 9%Ni in the austenite matrix.

Reactor noise analysis of the experimental fast reactor JOYO

Experimental fast reactor JOYO achieved its first criticality in 1977 spring. Since then a series of reactor noise analysis has been carried out. During zero power testing, the reactor kinetic parameters, i.e. β/l and subcriticality, were obtained, applying band-pass filter and polarity correlation methods. As the high power operation started from 1978, the reactor noise analysis based on neutron flux spectra has been proceeded. This paper presents study of noise analysis for rather lower frequency region at high power operation. At 50 Mwt power level operation, noises of neutron flux signals, reactor inlet temperatures and subassembly outlet temperatures were measured, and then auto power spectral densities and coherence functions were calculated. The frequency range for the noise analysis is for 10 −3 − 10 −1 Hz. In the APSD of neutron flux, a low peak was observed at 2.5×10 −2 Hz. For neutron noise, it was estimated that the noises in the frequency region lower than 1.5 × 10 −2 Hz is a core inherent temperature noises, while that in the frequency region higher than 1.5 × 10 −2 Hz is a core inherent neutron noise. The coherence function of neutron fluxes, which are signals of ex-core detectors located at opposite positions against the core center, indicated that some space dependent phenomena of neutron noise might exist in 5×10 −3 − 6×10 −2 Hz. The space dependency in 5×10 −3 − 1.5×10 −2 Hz were estimated to be produced by the sodium temperature noise due to the insufficient mixing of two primary coolant loops. In 1.5×10 −2 − 6×10 −2 Hz, the peak existed at 2.5×10 −2 Hz and reverse phases were observed in the coherence of two neutron signals at opposite side of the core. With the investigation of phase relations between neutron flux and subassembly outlet temperature in this frequency region, the phenomenon was estimated to be due to mechanical vibration of some reactor core components. When the static gain of reactivity-to-power transfer function increased, the changes of reactor noise characteristics were observed. Since the change of the gain indicates a change of the reactor core characteristics, the observation of reactor noise characteristics has a potential for the surveillance of the reactor core.

On-line diagnosis algorithm based on noise analysis

A minicomputer-based on-line diagnosis algorithm was developed and its verification test was performed. This algorithm identifies the plant status by applying a statistical pattern recognition technique to various multidimensional noise pattern vectors. In case the reference noise patterns are prepared for such abnormal plant states that the noise response characteristics are linear to different abnormality severeness values, this algorithm can discriminate the category and relative severeness of abnormality, irrespective of its severeness. The test was performed on-line with a simple analog simulator which was modeled on the experimental fast reactor JOYO. Seven pattern vectors were used to diagnose three categories of abnormality including abnormal control rod vibration. Reference patterns were learned under only one degree of severeness for each abnormality category. Test results showed the effectiveness of the proposed algorithm. This indicates that the number of reference patterns to be prepared can be significantly decreased.