The Post-Irradiation Examination (PIE) data are quite useful to validate the evaluated nuclear data. Recently, a new experimental program, REGAL (the Rod-Extremity and Gadolinia AnaLysis), has been ...proposed and launched. During this program, the PIE data are acquired for the Gd-bearing fuel rods during (or after) the initial Gd burnout. These PIE data would be different from those for the normal UOX or MOX fuel rods from a viewpoint of nuclide transmutation process. In the present work, we attempt to quantify this difference through calculating sensitivity coefficients of nuclide number densities of the Gd-bearing rods using the depletion perturbation theory and performing uncertainty quantification calculations with the sensitivity coefficients and the nuclear data covariance data. Through the numerical analyses, we found that the impact of the uncertainties of the nuclear data of Gd isotopes on the number densities of actinoids is negligible. The nuclear data-induced uncertainties of Gd isotopes number densities are relatively large during the Gd depletion, and those become small after the Gd burnout. After the Gd burnout, the relative standard deviations of Gd-155 and -157 number densities are approximately 6% and 13%, respectively, and the nuclear data of the parent nuclides of these Gd isotopes, Gd-154 and -156, are also important as well as those of Gd-155 and -157.
Sensitivities of
k
∞
and nuclides number densities during nuclear fuel burnup with respect to nuclear data are calculated with a reactor physics code system CBZ. Sensitivity calculations are carried ...out with the depletion perturbation theory applicable to nuclear fuel assemblies including burnable absorbers. Numerical results are presented both for BWR and PWR assemblies, and those demonstrate usefulness and effectiveness of burnup sensitivity calculation capabilities for LWR fuel assemblies.
CBZ is a general-purpose reactor physics analysis code system, and FRBurner, which focuses on fast reactor burnup calculations, was developed recently with diverse combinations of available ...methodologies. Verification of this module is conducted with the OECD/NEA fast rector benchmark since this benchmark provides various types of fast reactors. Four key reactor physics parameters, effective neutron multiplication factor k
eff
, effective delayed neutron fraction β
eff
, sodium void reactivity ∆ρ
void
, and Doppler reactivity ∆ρ
Doppler
are the focus and compared to two references provided by JAEA and CEA, respectively. The biases between the results from FRBurner and the JAEA and CEA references on each of the above key parameters are less than 0.5%, 1%, 3% and 7%, and less than 1.0%, 4%, 12%, and 12%, respectively. The comparison indicates that the FRBurner module would provide acceptable results for general-type fast reactor physics analysis in research. As one innovation, the detailed burnup chain model, which is significantly different from a generally used pseudo fission product model in fast reactor neutronic analysis, is applied in FRBurner. The detailed burnup chain model helps FRBurner explicitly provide information about the inventory of fission products for nuclear waste management and spent fuel reprocessing.
Useful and valuable measurement data obtained at the solid-moderated core for the development of accelerator-driven systems (ADSs) have been accumulated at the Kyoto University Critical Assembly ...(KUCA), and some of them have been open to the public. In order to efficiently utilize these data, experimental analyses with deterministic calculation procedures are helpful. In the present manuscript, a numerical benchmark problem is established. This benchmark problem can be utilized by users of the ADS-related measurement data obtained at the KUCA A-core to verify their own numerical tools devoted to experimental analyses. Material and geometrical specifications with reference solutions obtained by a continuous-energy Monte Carlo code MVP-II are provided.
In addition, numerical results obtained by a deterministic code system CBZ are also presented as an example. Through careful investigation about discretization on space and angle, guideline for proper discretization is provided. The CBZ results tend to underestimate the reference Monte Carlo solutions about 0.5%∆k/kk', and calculations of simplified core models suggest that this is caused by neutron leakage treatment in finite systems or resonance self-shielding treatment in CBZ.
The quick and accurate neutron and photon transport calculations are desired in the optimization calculations for the neutron source design, and in the present work, we establish the deterministic ...neutron and photon transport calculation procedure with the nuclear reactor physics calculation code system CBZ. Numerical calculation conditions are carefully chosen, and the efficient and practical condition is determined. Test calculations are carried out for the simple cylindrical systems with various kinds of neutron moderators, and the results are compared with the reference solutions obtained by the continuous-energy Monte Carlo code PHITS. Generally good agreements are obtained for all the benchmark problems. In addition, another problem with the detailed geometry for the neutron source is prepared. In this realistic problem also, good agreement is obtained between CBZ and PHITS. These results demonstrate the high accuracy of CBZ in the application to the design optimization calculations for the neutron source.
•A deterministic transport code CBZ was modified for BNCT irradiation field study.•The improved CBZ code can obtain almost the same result as the PHITS code.•The calculation time is about 2,000 times faster than the PHITS code.•The modified CBZ code is useful for iterative optimization of irradiation field.
There are several different types of the integral data of nuclear fission systems, and the accuracy of the numerical predictions of them are related with each other via the nuclear data commonly used ...in the numerical simulations. Thus, it is sometimes possible that measurement data of one type of the integral data are used to improve the prediction accuracy of the different type of the integral data. To quantitatively evaluate such possibility, the similarity of the different types of the integral data is important. In this study, we quantitatively evaluate the similarity between neutron multiplication factors and nuclide inventories during nuclear fuel burnup from the viewpoint of the nuclear data uncertainties using the representativity factors. Using the Burner module of the CBZ reactor physics code system for fuel pin-cell burnup problems, we calculated the sensitivity of the neutron infinite multiplication factors and the inventories of the 17 actinoids at several fuel burnup points. The neutron multiplication factor during the fuel burnup was considered the target parameter for which the prediction accuracy is improved, and the degree of similarity of the nuclide inventory data during burnup to the target was quantitatively evaluated using the representative factor. Subsequently, multiple nuclide inventory data were combined using the concept of the extended bias factor method to create a fictitious parameter, and we investigated how much this fictitious parameter can increase the representativity factor for the target parameter. As a result, the representativity factor for the target parameters could be increased to more than 0.8 using some nuclide inventory data, and up to 0.92 depending on the burnup, even taking into consideration the measurement error.
Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous ...change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, universities, and corporations, focusing on the areas in reactor physics and thermal-hydraulics since the Fukushima accident to the present date.
•Sensitivities of peak reactivity of LWR fuel assemblies are quantified.•Usefulness of the depletion perturbation theory-based capability is demonstrated.•Peak reactivity uncertainty induced by Gd ...(n,g) cross section is generally small.•Gd-155 nuclear data uncertainty is rather important than Gd-157 nuclear data uncertainty.
The reactivity of a fuel assembly including burnable absorbers can become the largest at low fuel burnup, so the accuracy of reactivity calculations for such systems is important. To investigate this issue, the impact of gadolinium isotopes’ nuclear data on neutron multiplication factor k is quantified. Sensitivities of k to nuclear data are calculated from 0 to 20 GWD/t for a BWR 3 × 3 multicell model. Sensitivity to gadolinium-157 (n,γ) cross section becomes the largest at the zero burnup. Sensitivity to gadolinium-155 (n,γ) cross sections takes the two largest values and the second one is observed around fuel burnup where the reactivity reaches its peak. Sensitivities are also calculated for BWR and PWR assemblies, and similar trends are observed. Finally, nuclear data-induced uncertainties of k are quantified. Gadolinium-157 contribution is the largest at zero burnup, and gadolinium-155 contribution is relatively important around fuel burnup corresponding to the reactivity peak.
In experimental benchmarks of the accelerator-driven system (ADS) conducted at the Kyoto University Critical Assembly (KUCA), the prompt neutron decay constant
was measured using two types of pulsed ...neutron sources, i.e. a D-T neutron source and a spallation neutron source driven by a 100-MeV proton beam. The measurement results of
are useful information to validate the numerical results predicted by the prompt
-eigenvalue calculation. In this study, the numerical analysis of
using a multi-energy group S
N
neutron transport code was carried out for the uranium-lead zoned experimental cores. To reduce the discretization error owing to the deterministic code, the KUCA geometry was modelled in detail as a three-dimensional heterogeneous plate-by-plate geometry, and an improved variant of EO
N
quadrature was utilized. In addition, the sensitivity coefficients of
with respect to nuclear data were efficiently evaluated by first-order perturbation theory, followed by nuclear data-induced uncertainty quantification based on the 56 neutron-energy group SCALE covariance library. Consequently, the numerical results of
were validated successfully by the experimental results of the pulsed neutron source method, compared with the range of the nuclear data-induced uncertainties.
This paper presents the results of validation calculations of the nuclear fuel depletion calculation module of the CBZ reactor physics code system, CBZ/Burner. Validation calculations were conducted ...using the post irradiation examination data obtained at Fukushima-Daini Unit 2 and at Takahama Unit 3. The nuclide number densities calculated with CBZ/Burner were compared with the measurement values, and generally good agreement was obtained. The sensitivity coefficients of the nuclide number densities with respect to nuclear data were calculated for all concerned nuclides with the depletion perturbation calculation capability of CBZ/Burner, and the nuclear data-induced uncertainties of the nuclide number densities were quantified. From the numerical results, we can conclude that the nuclear fuel depletion calculation module for LWR in the CBZ code system was successfully validated.
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