Experimental Fast Reactor JOYO Research Articles

Status of the 226Ra nuclear data library and its impact on the production amount of 225Ac via the 226Ra (n,2n) reaction

ABSTRACT 225Ac is a promising alpha emitter for targeted alpha therapy. However, the current capability of 225Ac supply is limited to approximately 63 GBq/y, primarily relying on the natural source of 229Th stocked at a few institutes. Therefore, alternative 225Ac production methods are highly desired. The research and development of 225Ac and its parent nuclide production methods using accelerators and reactors to target 226Ra and 232Th are actively pursued worldwide. Hence, the authors focused on the 226Ra(n,2n)225Ra method using fast neutron spectra as an application of the experimental fast reactor Joyo. This study investigated the status of nuclear data libraries for 226Ra, an essential target for 225Ac production, and evaluated the impact of different nuclear data libraries on the amount of 225Ac produced. Consequently, cross-sections with covariance data were stored in TENDL-2021, JEFF-3.3, and EAF-2010 but not in ENDF or JENDL, the major nuclear data libraries. Furthermore, no consistency occurred among the respective nuclear data, and the 225Ra production amount varied.

Evaluation of the production amount of 225Ac and its uncertainty through the 226Ra(n,2n) reaction in the experimental fast reactor Joyo

ABSTRACT 225Ac is a promising nuclide for targeted alpha therapy. At present, the nuclide is mainly extracted from 229Th generators, but its total radioactivity of the world is only 63 GBq/y. Therefore, it is unclear whether this method alone can meet future global medical demand. In this study, we analyzed the production amount of 225Ac and its uncertainty due to the neutron flux and nuclear reaction cross-sections while analyzing the generation process of 225Ac using the experimental fast reactor Joyo operated by the Japan Atomic Energy Agency (JAEA). The analysis showed that irradiating 1 g of 226Ra for 45 d at the core center of Joyo can produce 15.4 ± 6.2 GBq of 225Ac. Therefore, with six irradiations per year (45 d × 6 = 270 d/year), Joyo can produce 55.3 to 129.1 GBq of 225Ac, equivalent to 86 to 200% of the current world supply.

Tensile properties of modified 316 stainless steel (PNC316) after neutron irradiation over 100 dpa

ABSTRACT The effects of fast neutron irradiation on tensile properties of 20% cold-worked modified 316 austenitic stainless steel (PNC316), which was improved swelling resistance and creep strength, claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400°C and 735°C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening depending on irradiation temperature. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400°C and 500°C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.

Evaluation of transmutation rate of some LLFP in experimental fast reactor JOYO

A transmutation process of three long-lived fission products (79Se, 99Tc and 107Pd) in the experimental fast reactor JOYO is postulated. The possibility of increasing the transmutation rate utilizing the high neutron flux present in the JOYO reactor by loading neutron-moderating subassemblies in the reflector zone has been investigated. A cluster of reflector subassemblies was replaced with beryllium or zirconium hydride (ZrH1.65) moderated subassemblies. These moderated subassemblies surrounded one central test subassembly that would contain the three long-lived fission products (LLFP) simultaneous and without isotopic separation. ChainSolver 2.34 code is used to calculate the transmutation rates. In this study, the new characteristics of LLFP transmutation in a fast reactor using moderator materials were shown for future applications.

Neutronic Study on Ac-225 Production for Cancer Therapy by (n,2n) Reaction of Ra-226 or Th-230 Using Fast Reactor Joyo

Ac-225 has lately drawn considerable attention as a radioisotope for targeted alpha therapy treatment for certain types of prostate, blood-derived, and disseminated cancers, but its supply is limited. Therefore, we investigated the production method of Ac-225 by nuclear transmutation in a fast neutron reactor. The authors investigated irradiation of Ra-226 or Th-230 as a target nuclide in the experimental fast reactor Joyo, owned and operated by Japan Atomic Energy Agency, which has abundant fast neutrons and a large loading region with high heat removal capacity. Ra-226 is in increasing demand as a target nuclide to produce Ac-225. Therefore, as another option, we selected Th-230, which is 50 times more abundant than Ra-226 in natural uranium, as an alternative nuclide. Irradiation of Ra-226 and Th-230 with high energy neutrons above the threshold causes an (n,2n) reaction, producing Ra-225 and Th-229, respectively, which are the parent nuclides of Ac-225. The analyses showed that 47 GBq of Ac-225 can be generated annually by irradiating 1 g of Ra-226, and 6.5 GBq of Ac-225 can be semi-permanently generated every year by one-time irradiation of 50 g of Th-230 for 10 years (5 EFPY). It can be concluded that 100 MWt Joyo has potential to produce more than 70% of the current global supply of Ac-225 and/or to generate the parent nuclide Th-229, which keeps producing Ac-225 for thousands of years.

Design of Cesium target assembly with YH2 moderator and Gd thermal neutron shielding to produce 135Cs for LLFP transmutation study using the experimental fast reactor Joyo

In this design study, we investigated the feasibility of producing a sufficient amount of 135Cs long-lived fission product (LLFP) in the Joyo fast reactor by irradiation of natural cesium (133Cs). The irradiation is expected to contribute to providing 135Cs sample for transmutation cross-section measurements for enhancing the accuracy of the LLFP nuclear data. Two irradiation fields, one located in the outer reflector and another one in the inner reflector which was adjacent to the core were investigated, and appropriate target subassemblies were designed by using the MCNP and FISPACT-II codes. In the inner reflector case, the YH2 moderator material was adopted for the neutron spectrum adjustment within the target subassembly. The gadolinia thermal neutron shielding is proposed to suppress the power peak of the adjacent fuel assembly. The result confirmed that a sufficient amount of 135Cs can be produced at 1200 Effective Full Power Day (EFPD) in Joyo reactor (Mark IV core, 100 MWth).

Development of fuel performance analysis code, BISON for MOX, named Okami: Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

• 2-D analyses with modelling for MOX fuel restructuring in use of fast reactor. • Effects of O/M ratio and pellet-cladding gap width on pore migration behavior. • Irradiation behavior evaluation for the MA-bearing MOX irradiation test in Joyo. • Influence of pellet eccentricity on central void formation in 2-D analyses. Fuel composition, i.e. Pu content, minor actinide (MA) content, oxygen-to-metal (O/M) ratio, etc., affects the irradiation behavior of MA-bearing mixed oxide (MOX) for fast reactor. To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with Idaho National Laboratory (INL) and have installed a pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation in the experimental fast reactor Joyo was evaluated by a two-dimensional (2-D) analysis. To solve partial differential equations for energy conservation and advection diffusion equations to calculate temperature and pore migration, respectively, a correlation of thermal conductivity as a function of porosity, temperature, deviation from stoichiometry, Am content and Np content was used. In addition, the pore migration velocity was calculated by using a correlation depending on vapor pressure applying Rand-Markin's model. Four MA-bearing MOX test pins with different O/M ratio and pellet/cladding gap size were irradiated in B14 test in Joyo to study the effects of O/M ratio and gap size on irradiation behavior at the beginning of irradiation. In post-irradiation examinations (PIE), a remarkable restructuring was observed in stoichiometric fuels, and a discrepancy in central void diameter among various O/M ratios. This restructuring behavior could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void is assumed to offset from the geometric center to the larger-gap side due to the pellet eccentricity, but the PIE results were inconsistent in this regard. It is expected that the irradiated pellet might displace to another side during sample preparation in PIE from a reverse offset location during irradiation by comparing the result of 2-D analysis and radial profile of fuel density.

Release behavior of radionuclides from MOX fuels irradiated in a fast reactor during heating tests

In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FRs were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.

Restructure behavior analysis in fast breeder reactor MOX fuel by X-ray CT

ABSTRACT Both high- and low-density MOX fuel pellets of uranium and plutonium oxides were irradiated in the experimental fast reactor JOYO. After irradiation, these fuel pellets were examined by X-ray CT and their irradiation behavior was evaluated for formation of the central void. In particular, the central void size and temperature of fuel pellets at the beginning and end of irradiation were analyzed. The central voids in the low-density fuel pellets were bigger than those of the high-density fuel pellets at the same linear heating rate (LHR), and the threshold LHR and temperature at which the central voids were formed were lower than those of the high-density fuel pellets. It was understood from these results that the irradiation behaviors of high- and low-density fuel pellets were different.

Neutron irradiation effects on the microstructural development of tungsten and tungsten alloys

Data on the microstructural development of tungsten (W) and tungsten rhenium (Re) alloys were obtained after neutron irradiation at 400–800 °C in the Japan Materials Testing Reactor (JMTR), the experimental fast test reactor Joyo, and the High Flux Isotope Reactor (HFIR) for irradiation damage levels in the range of 0.09–1.54 displacement per atom (dpa). Microstructural observations showed that a small amount of Re (3–5%) in W–Re alloys is effective in suppressing void formation. In W–Re alloys with Re concentrations greater than 10%, acicular precipitates are the primary structural defects. In the HFIR-irradiated specimen, in which a large amount of Re was expected to be produced by the nuclear transmutation of W to Re because of the reactor's high thermal neutron flux, voids were not observed even in pure W. The synergistic effects of displacement damage and solid transmutation elements on microstructural development are discussed, and the microstructural development of tungsten materials utilized in fusion reactors is predicted.

Problem on MATXS Files with Multiple Temperature Cross Section Data

Abstract This paper describes a problem on multigroup MATXS files with multiple temperature cross section data. This problem was newly found out through nuclear analyses for a simple fast reactor model with MATXSLIB-J40, which is a multigroup cross section data file (processed temperature: 300, 600, 900, 1200, 1800 K) of the latest Japanese Nuclear Data Library version 4.0 (JENDL-4.0) processed with the NJOY99 code. Neutron and gamma fluxes inside a simple cylindrical model (40 cm in radius and 50 cm in height; homogenized mixture of U/Pu fuel, stainless steel and sodium; two cases of 300 K and 600 K in temperature) for the experimental fast reactor JOYO in Japan Atomic Energy Agency were calculated with the DORT code and a multigroup library generated from MATXSLIB-J40 using the TRANSX2 code. Then the total neutron fluxes were almost the same at 300 K and 600 K, while the total gamma fluxes at 600 K were by 10% higher than those at 300 K. Reasons of this discrepancy for the total gamma fluxes were investigated in detail. As a result, it was found out that the scattering matrix data from neutron groups to gamma groups in the multigroup library at 600 K were not always correct, because the MATXS data format processed with NJOY was not consistent with that defined in TRANSX for gamma production data (10d part in MATXS) in temperatures except for the first temperature. Note that this problem can also appear in MATXS files with multiple temperature cross section data of other nuclear data libraries. A simple program for modifying MATXS files to ones suitable to TRANSX was made in order to solve this problem. MATXSLIB-J40 will be revised with this program.

Fuel behavior analysis code FEMAXI-FBR development and validation for core disruptive accident

Japan Nuclear Energy Safety Organization (JNES) is a Technical Safety Organization (TSO) for Japanese Nuclear Regulation Authority to provide technical information on safety reviews including incident and accident analyses of nuclear power plants including Fast Breeder Reactor (FBR). JNES has developed an ASTERIA-FBR code system as an integrated Core Disruptive Accident (CDA) analysis code of the regulatory side for the purpose of evaluating the Unprotected Loss-of-Flow (ULOF) accident of FBR cores. The ASTERIA-FBR consists of three main modules, a thermal-hydraulics calculation module; CONCORD, a fuel behavior calculation module; FEMAXI-FBR, and a neutronic calculation module; dynamic GMVP. The FEMAXI-FBR is based on a LWR fuel behavior simulation code FEMAXI-6 and modified the material properties, the calculation models for steady state and transient operational conditions. The FEMAXI-FBR has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient test data. Furthermore, the fuel temperature calculation model was validated by B5D Power-To-Melt (PTM) test at the experimental fast reactor JOYO in Japan. Through these validation, the code is found to be capable of evaluating FBR fuel pin behavior from normal to accidental conditions precisely.

Void denuded zone formation for Fe–15Cr–15Ni steel and PNC316 stainless steel under neutron and electron irradiations

Irradiation-induced void denuded zone (VDZ) formation near grain boundaries was studied to clarify the effects of minor alloying elements on vacancy diffusivity during irradiation in practical PNC316 stainless steel developed for nuclear reactor core materials. The test materials were Fe–15Cr–15Ni steel without additives and PNC316 stainless steel; the latter contains minor alloying elements to improve the void swelling resistance. These steels were neutron-irradiated in the experimental fast reactor JOYO at temperatures from 749K to 775K and fast neutron doses of 18–103dpa, and electron irradiation was also carried out using 1MeV high voltage electron microscopy at temperatures of 723K and 773K and doses up to 14.4dpa. VDZ formation was analyzed by TEM microstructural observation after irradiation by considering radiation-induced segregation near the grain boundaries.VDZs were formed near random grain boundaries with higher misfit angles in both Fe–15Cr–15Ni and PNC316 steels. The VDZ widths in the PNC316 stainless steel were narrower than those for the Fe–15Cr–15Ni steel for all neutron and electron irradiations. The VDZ width analysis implied that the vacancy diffusivity was reduced in PNC316 stainless steel as a result of interaction of vacancies with minor alloying elements.

Evaluation of Burnup Reactivity Coefficients Measured in Experimental Fast Reactor JOYO MK-I Duty Power Operation Cycles

To provide a reactor physics benchmark for burnup reactivity coefficients, experimental data, showing the relationship between excess reactivity and accumulated thermal power acquired during the experimental fast reactor JOYO MK-I duty power operation in the late 1970s, have been evaluated and analyzed. To improve the prediction accuracy of nuclear characteristics through the use of integral experimental data, nominal values and uncertainties, including correlations of the experimental data, were evaluated. All possible uncertainty factors were evaluated and quantified by utilizing knowledge obtained after the MK-I duty power operation and calculation results based on the latest reactor physics analysis methods. Meanwhile, the present evaluated data have been reviewed and approved by the International Reactor Physics Experiment Evaluation Project, with the expectation that these data will be widely used. In the present paper, the evaluation of nominal values and uncertainties is described with a focus on the measurement technique uncertainty, which is a dominant uncertainty factor of the burnup reactivity coefficient. In addition, new analysis results of the benchmark problem are shown by the use of the latest Japanese evaluated nuclear data JENDL-4.0.

Development of neutron spectrum unfolding method for advanced nuclear emulsion

In order to realize neutron spectroscopy in high-intensity gamma ray fields such as those present in spent nuclear fuel storage facilities using an advanced nuclear emulsion based on a non-sensitized OPERA film with AgBr grain sizes of 60 nm, a new neutron spectrum unfolding method has been developed. The response functions were evaluated in the neutron energy range between 0.1 and 4.5 MeV by Monte Carlo calculations. To realize a highly reliable unfolding method that does not require an initial guess spectrum, an algorithm was formulized based on the maximum entropy principle and the maximum likelihood method, and the new unfolding code MEALU was developed. The code performance was checked by analyses of mock-up data. As an example, the neutron spectrum from fresh fuel of the experimental fast reactor Joyo was calculated and track lengths were simulated using the calculated response function and the estimated spectrum. Comparison between the original neutron spectrum and the unfolded one confirmed the effectiveness of the proposed method.

Application of statistical method for FBR plant transient computation

It is obvious that design tolerances, errors included in operation, and statistical errors in empirical correlations effect on the transient behavior. The purpose of the present study is to apply above mentioned statistical errors to a plant system computation in order to evaluate the statistical distribution contained in the transient evolution. A selected computation case is the turbine trip test conducted at 40% electric power of the prototype fast reactor “Monju”. All of the heat transport systems of “Monju” are modeled with the NETFLOW++ system code which has been validated using the plant transient tests of the experimental fast reactor Joyo, and “Monju”. The effects of parameters on upper plenum temperature are confirmed by sensitivity analyses, and dominant parameters are chosen. The statistical errors are applied to each computation deck by using a pseudorandom number and the Monte-Carlo method. The dSFMT (Double precision SIMD-oriented Fast Mersenne Twister) that is developed version of Mersenne Twister (MT), is adopted as the pseudorandom number generator. In the present study, uniform random numbers are generated by dSFMT, and these random numbers are transformed to the normal distribution by the Box–Muller method. Ten thousands of different computations are performed at once. In every computation case, the steady calculation is performed for 12,000s, and transient calculation is performed for 4000s. In the purpose of the present statistical computation, it is important that the base regions of distribution functions should be calculated precisely. A large number of computations are performed in order not to reduce the precision of the base regions of the distributions. Through a set of computation, the effect of errors on plant transient is obtained as the statistical distribution. In addition, the reduction method of trial numbers is considered. And results using the method express base regions of the distribution.

Microstructure development of dispersion-strengthened tungsten due to neutron irradiation

This work investigated the effects of neutron irradiation on microstructure development in pure and dispersion-strengthened W (i.e., lanthanum (La)-doped W and potassium (K)-doped W). The irradiation experiments were performed in the experimental fast reactor Joyo. The irradiation temperature and damage were in the range 531–756°C and 0.42–0.47dpa, respectively. Voids, dislocation loops, and black dots were observed in pure, La-doped, and K-doped W after neutron irradiation with almost identical size and number density of the defect clusters. Likewise, the hardening of pure, La-doped, and K-doped W because of neutron irradiation was nearly identical. Thus, the dispersion of the second phase in La-doped and K-doped W did not affect the microstructure development and irradiation hardening in tungsten under neutron-irradiation conditions in this work.

「常陽」における確率論的安全評価(その5) - 炉外事象、被ばく評価、PSAの結論 -

「常陽」における確率論的安全評価(その5) - 炉外事象、被ばく評価、PSAの結論 -

Property change of advanced tungsten alloys due to neutron irradiation

This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073K and 0.15–0.47dpa, respectively. TEM images of all samples after 0.42dpa irradiation at 1023K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening.

Restructuring and redistribution of actinides in Am-MOX fuel during the first 24h of irradiation

Abstract In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95.