Kiyoshi Sakurai

Measurement and evaluation of neutron spectra above 0.1 MeV in the JMTR

Abstract The evaluation of fast neutron spectra from the Japan Materials Testing Reactor (JMTR) have been performed by using the critical facility of the JMTR and by a combination of the multi-foil activation method and the adjustment codes (SAND II and NEUPAC). In order to measure and evaluate the neutron spectra above 0.1 MeV, resonance detectors such as manganese, gold and copper have been used to determine the neutron flux level in the 1 E region and threshold detectors such as silver, rhodium, indium, uranium, aluminum, magnesium and titanium have been used to determine the neutron flux level above 0.1 MeV. The foils for the measurement of the neutron reaction rate were separately irradiated. The 115In(n,n′)115mIn reaction is used for the monitoring of the average fast neutron flux in the irradiation period, and the slight difference of each irradiation condition was corrected. The guess spectra for the neutron spectrum adjustment were calculated by using the one-dimensional discrete-ordinates code ANISN with the slab model for the JMTR core. Some important points were concluded through the adjustment procedure of the neutron spectrum: the adjusted spectrum from 0.1 to 1 MeV depends on the accuracy of the neutron cross section data for the threshold detectors such as silver and rhodium, and also on the accuracy of these reaction rates. The ratios of neutron flux above 0.183 MeV to neutron flux above 1 MeV were calculated from the guess spectra and the adjusted spectra, and the ratios were in good agreement with each other.

Experimental evaluation of fission spectrum averaged cross sections of 93Nb(n, n′)93mNb and 199Hg(n, n′)199mHg reactions

Abstract The fission spectrum averaged cross section of the 93 Nb(n, n′) 93m Nb reaction has been performed to obtain (125 ± 28) mb. The value is about 15% larger than that calculated from the Hegedus cross section data. For the 199 Hg(n, n′) 199m Hg reaction, (252 ± 20) mb has been obtained.

Use of new threshold detectors 93Nb(n, n′)93mNb and 199Hg(n, n′)199mHg for neutron spectrum unfolding

Abstract The feasibility of using new threshold detectors 39 Nb(n, n′) 93m Nb and 199 Hg(n, n′) 199m Hg was examined. The neutron spectrum YAYOI glory-hole was unfolded with 13 reaction rates including the 93 Nb(n, n′) 93m Nb or the 199 Hg(n, n′) 199m Hg reaction rate characteristic neutron flux density values (total, above 1 and 0.1 MeV) were calculated from the best estimate for the input spectrum and from the unfolded neutron spectrum. The characteristic neutron flux density values calculated from the input spectrum were about 20% smaller than those calculated from the output neutron spectrum. The neutron flux density values calculated from the neutron spectrum unfolded with 13 reaction rates including the 93 Nb(n, n′) 93m Nb reaction rate were about 5 to 10% higher than those rate, the characteristic neutron flux density values were about 1 to 3% smaller than those calculated from the neutron spectrum unfolded without the 199 Hg(n, n′) 199m Hg reaction rate. These new threshold detectors were successfully used for neutron spectrum unfolding with the SAND II code.

Reactivity Measurements of Erbium at Tank-type Critical Assembly and Analyses Using ENDF/B-VI, JEF-2.2 and preliminary JENDL-3.3 Libraries

Reactivity effects due to neutron absorption of Erbium were measured at the Tank-type critical Assembly (TCA) for validation of Erbium’s nuclear data. Absorption effects in the epithermal energy range were measured as well as in the thermal energy range. The reactivities obtained by the integral experimental method were calculated by a two-dimensional discrete-ordinates code using ENDF/B-VI, JEF-2.2 and preliminary JENDL-3.3. All these libraries provides approximately 10% higher reactivities than the measured values in the thermal energy range.

Estimation of Subcriticality with the Computed Values Analysis using MCNP of Experiment on Coupled Cores.

Experiments on coupled cores performed at TCA were analysed using continuous energy Monte Carlo calculation code MCNP 4 A. Errors of neutron multiplication factors are evaluated using Indirect Bias Estimation Method proposed by authors. Calculation for simulation of pulsed neutron method was performed for 17×17+5G+17×17 core system and its of exponential experiment method was also performed for 16×9+3G+16×9 and 16×9+5G+16×9 core systems. Errors of neutron multiplication factors are estimated to be -1.5-0.6% evaluated by Indirect Bias Estimation Method. Its errors evaluated current pulsed neutron method and exponential experiment method are estimated to be 7%, but it is below 1% for estimation of subcriticality with the computed values applied Indirect Bias Estimation Method. Feasibility of subcriticality management is higher by application of the method to full scalefuel strage facility.

Numerical validation of a modified neutron source multiplication method using a calculated eigenvalue

Abstract To overcome the difficulties of the conventional neutron source multiplication method, the authors have developed the Indirect Bias Estimation Method. This method obtains the bias in calculated K eff using the difference between measured and calculated neutron count rates. This bias in calculated K eff is used for an adjustment of the calculated K eff , deriving a truer K eff . Using neutron diffusion calculations, numerical experiments simulating the neutron source multiplication method were performed to validate this method. It could be shown how closely a true K eff can be reproduced using this method. It was found that an accurate estimation of subcriticality requires neutron count rate measurements at more than three locations.

Accurate Estimation of Subcriticality Using Indirect Bias Estimation Method, (II) Applications

Subcriticalities were estimated by applying the Indirect Bias Estimation Method to subcritical experiments on a light-water moderated/reflected low-enriched UO 2 lattice cores. Two measurable values, prompt neutron time-decay constant and spatial-decay constant were calculated using MCNP 4A and JENDL-3.2. With these calculation errors, the biases in calculated reactivity were derived from the Indirect Bias Estimation Method. The differences between the calculated and measured spatial-decay constants were more or less at the same extent of experimental errors. These results show that the accuracy of subcriticality estimation of MCNP 4A and JENDL-3.2 ranges within the uncertainty which can be achieved by the exponential experiment. The differences between calculated and measured prompt neutron decay constants derive significant biases in calculated reactivity. The subcriticalities were estimated by using the effective multiplication factors adjusted based on these biases in calculated reactivity.

Measurements and Analyses of Reactivity Effect of Fission Product Nuclides in Epithermal Energy Range

Experimental data usable for evaluating cross sections of main fission product elements (Rh, Cs, Nd, Sm, Eu and Gd) in the epithermal energy range were measured. A cadmium-covered vessel containing a pure water or an aqueous solution of a fission product element was inserted at the center of TCA (Tank-type Critical Assembly) core. Reactivity effects were obtained by the difference in the critical water levels between a pure water and an aqueous solution in the vessel. The measured reactivity was more than 1 φ and it was greater than the experimental uncertainties. Since the adjoint thermal flux below the cadmium-cutoff energy are largely depressed in the vessel, the reactivity effects in epithermal energy range could be measured. The analyses for the experiments were performed using the SRAC code system and neutron transport calculation code TWOTRAN. The exact Perturbation theory was applied to calculate the reactivity effects of fission product elements. The calculated reactivity effects using JENDL-3.2 ...