Kurosawa, Ryohei; Okada, Shota; Sakai, Akihiro; Nakata, Hisakazu; Amazawa, Hiroya
JAEA-Data/Code 2015-005, 82 Pages, 2015/06
The calculation tool of neutron flux at materials within and around the research reactor was developed so that the user more easily evaluate radioactivity inventory in radioactive waste generated from the decommissioning of research reactors at various conditions. The tool consists of some computer programs which calculate macroscopic effective cross section at materials, calculate the neutron flux at materials within and around the research reactor, and edit the neutron flux to evaluate the radioactive inventory. This report describes the outline of evaluation method of neutron flux at materials within and around the research reactor, the structure and functions of the calculation tool of neutron flux, input and output data, and sample run with the tool.
; ; ; Iguchi, Yukihiro; ;
JNC-TN3410 2000-014, 43 Pages, 2000/09
; Sato, Wakaei*;
JNC-TN9400 2000-037, 87 Pages, 2000/03
ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and -value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...
JNC-TJ9450 2000-002, 112 Pages, 2000/03
This report is intended to make it easier to apply the measured data obtained from the Gap Streaming Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about two months beginning at the start of March, 1992 as the sixth one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Gap Streaming Experiment was planned to obtain the data of neutron streaming characteristics in the inclosure system above the core of an advanced fast reactor for verification and improvement of the analysis method to be applied to the shielding design. A iron-lined solid or slit concrete assembly was placed, with or without a spectrum modifier forming soft incident neutron spectrum, behind the TSR-II reactor of Tower Shielding Facility. Inserting central cylinders and cylindrical sleeves gave various gap width and offset in the slit concrete assembly. Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured on radial traverse and on the axis behind the concrete assembly in almost all configurations. Neutron spectrum and fine radial distribution in high energy region was measured further in case of hard incident neutron spectrum, Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12140. "Measurements for the JASPER Program Gap Streaming Experiment"). Additional information reported by the assignee is utilized also.
Nakashima, Hiroshi; Masumura, Tomomi*; Tanaka, Susumu; Sakamoto, Yukio; Takada, Hiroshi; Tanaka, Shunichi; Nakane, Yoshihiro; Meigo, Shinichiro; Nakamura, Takashi*; Kurosawa, Tadahiro*; et al.
Journal of Nuclear Science and Technology, 37(Suppl.1), p.192 - 196, 2000/03
no abstracts in English
Nakashima, Hiroshi; Masumura, Tomomi*; Tanaka, Susumu; Sakamoto, Yukio; Tanaka, Shunichi; Nakane, Yoshihiro; Meigo, Shinichiro; Nakamura, Takashi*; Kurosawa, Tadahiro*; Hirayama, Hideo*; et al.
Journal of Nuclear Science and Technology, 37(Suppl.1), p.197 - 201, 2000/03
no abstracts in English
JNC-TJ9400 2000-009, 63 Pages, 2000/02
The present status of nuclear data for technetium (Tc)-99, which is a well-known fission product (FP), has been reviewed and investigated. And making use of the Kyoto university Lead Slowing-down Spectrometer (KULS), the cross section of the Tc (n, ) Tc reaction has been measured in the energy range from thermal to keV neutron energy with an Ar-gas proportinal counter. The neutron flux/spectrum has been monitored with a BF proportional counter, and the relative measurement has been normalized to the well-known standard capture cross section value for the Tc (n, ) Tc reaction at 0.0253 eV. Self-shielding corrections, especially near the resonance peaks, were made by the calculations with the MCNP code. Although the experimental data measured by Chou et al with a lead slowing-down spectrometer are higher in general, the energy dependency is similar to the present measurement. The evaluated data in ENDF/B-VI and JENDL-3.2 are higher near the resonances at 5.6 and 20 eV and above several 100 eV. A lead slowing-down spectrometer was installed coupled to a 46 MeV electron linac at the Research Reactor Institute, Kyoto university (KURRI). Characteristics of the Kyoto University Lead Slowing-down Spectrometer (KULS) were measured and (1)the relation between neutron slowing-down time t(s) and energy E(keV) (E=190/t in Bi hole and E=156/t in Pb hole) and (2)the energy resolution (40% in Bi and Pb holes) were experimentally investigated. (3)The neutron energy spectrum in the KULS was also measured by the neutron TOF method. The results obtained by the MCNP code were in general agreement with these experimental ones.
*; Kitada, Takanori*; Tagawa, Akihiro*; *; Takeda, Toshikazu*
JNC-TJ9400 2000-006, 272 Pages, 2000/02
Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by Khler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...
JAERI-Research 2000-003, p.110 - 0, 2000/02
no abstracts in English
Sato, Satoshi; Iida, Hiromasa; Plenteda, R.*; Valenza, D.*; Santoro, R. T.*
Fusion Engineering and Design, 47(4), p.425 - 435, 2000/01
no abstracts in English
JNC-TN9400 98-007, 147 Pages, 1998/11
A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under steady state and various transient operation modes. Heavy water is selected as a coolant material for heat removal of the SERAPH driver core during the experiments. Control rods are needed to conduct the experiments, and a control rod with heavy water follower is considered as one of the promising ideas and is now under investigation. In this idea, care must be taken to avoid production of local power peaks which are caused by neutron moderation in the follower and may appear in the vicinity of the boundary between the control rod and its neighboring fuel subassembly, since deuterium has an excellently high moderation power. Therefore, preparation of some methods of evaluating power density distribution in detail is required for control rod design. This report describes preparation of a set of neutronic calculation methods to evaluate intra-subassembly power density distribution including local power peaks around a control rod. A two-dimensional S transport calculation code TWOTRAN-II is selected as a tool for evaluating neutron transport phenomena near the control rod with no cares for statistical influence. A two-dimensional rectangular super-cell model, which is a physical model composed of a control rod and its surrounding fifteen fuel sub-assemblies, and a method to construct the super-cell model based on thirteen unit cells are created, considering neutron mean free path near a control rod. Two processing tools are newly developed to generate a material region map and mesh boundaries for an efficient super-cell construction procedure and to obtain pin-wise power densities based on calculated mesh-wise neutron flux data. The results in this report are expected to be ...
Suzudo, Tomoaki; Tuerkcan, E.*; H.Verhoef*
Nuclear Science and Engineering, 129(2), p.203 - 208, 1998/06
no abstracts in English
Takeda, Toshikazu*; Kitada, Takanori*; *; *
PNC-TJ9605 98-001, 267 Pages, 1998/03
As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method. A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. As for the nuclides to be analyzed, the elements of structure material, such as iron, nickel, chrome and sodium were considered. By the present method, all the reactions became larger at the deep region in the blanket. The maximum correction amounted as much as 5%. This tendency lessen the disagreement between the ordinary calculation and the experiment. It was made clear that the treatment in inter-band scattering term is veryimportant because it has large sensitivity on the result. An alternative method to determine the multiband parameters whieh method is based on more direct approach and is free from drawbacks in the present method, was also investigated. Part 2 : Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory. Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The continuous energy Monte-Carlo perturbation code has been developed by using not only the correlated sampling method which is already used before, but also the derivative operator sampling method. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. The change of eigenvalue caused by the change of sodium density in the GEM or dummy ...
PNC-TN9410 98-015, 81 Pages, 1998/02
The study was carried out within the framework of the PNC-CEA collaboration agreement. Data were provided, by CEA, for an experimental loading of a start-up core in Super-Phenix. This data was used at PNC to produce core flux snapshot calculations. CEA undertook a comparison of the PNC results with the equivalent calculations carried out by CEA, and also with experimental measurements from SPX. The resu1ts revealed a systematic radial flux tilt between the calculations and the reactor measurements, with the PNC tilts only 30-401 of those from CEA. CEA carried out an analysis of the component causes of the radial tilt. It was concluded that a major cause of radia1 tilt differences between the PNC and CEA calculations lay in the nuclear datasets used: JENDL-3.2 and CARNAVAL IV. For the final stage of the study, PNC undertook a sensitivity analysis, to examine the detailed differences between the two sets of nuclear data. The PNC flux calculations modelled SPX in both 2D (RZ) and 3D (hex-Z) geometries, using the diffusion programs CITATION and MOSES. The sensitivity analysis of the differences between the JENDL-3.2 and CARNAVAL IV nuclear datasets used the SAGEP calculational route. Both datasets were condensed to a single, non-standard, set of energy group boundaries. There were some incompatibilities in the cross-section formats of the two datasets. The sensitivity analysis showed that a relatively small number of nuclear data items contributed the bulk of the radial tilt difference between calculations with JENDL-3.2 and with CARNAVAL IV. A direct comparison between JENDL-3.2 and CARNAVAL IV data revealed the following. The Nu values showed little difference (<5|%). The only large fission cross-section differences were at low energy (<30% otherwise, with <10% typical). Although down-scattering reactions showed some large fractional differences, absolute differences were negligible compared with in-group scattering; for in-group scattering fractional ...
Yamada, Tadanori; ; Shirasu, Noriko
JAERI-Tech 97-072, 32 Pages, 1998/01
no abstracts in English
PNC-TN9410 97-088, 139 Pages, 1997/10
Critical experiments were carried out on Deuterium Critical Assembly (DCA) modification core. DCA modification core has two regions, that is, test region and driver region. The test region consists of various types of fuel and moderator, while the driver region remains the same as the original DCA core (ATR simulated core). Critical characteristics were measured with various types of core patterns and were compared with calculated values based on SCALE code system. Monte calro code KENO was found to be very accurate in the core analysis. The accuracy stays below 0.5 %dk/k in keff even if core configulation is extremely complicated.
Fletcher, J. K.
PNC-TN9410 97-065, 25 Pages, 1997/07
Takeda, Toshikazu*; *; Kitada, Takanori*; *
PNC-TJ9605 97-001, 100 Pages, 1997/03
This report is composed of the following two parts and appendix. (I)Improvement of the Method for Evaluating Reactivity Based on Monte Carlo Perturbation Theory (II)Improvement of Nodal Transport Method for 3-D Hexagonal Geometry (Appendix) Effective Cross Section of U Samples for Analyzing Doppler Reactivity in Fast Reactors Part I. Improvement of the Method for Evaluating Reactivity Bascd on Monte Carlo Perturbation Theory. Theoretical formulation in Monte Carlo perturbation method had been checked, and then introduced into a calculation code. The increase of CPU time is about 10 to 20 % compared to that if normal Monte Carlo code, in the cases of same number of history. This Monte Carlo perturbation method found to be effective, because results are almost reasonable and deviations of the results are especially small, by using the Monte Carlo perturbation code. However, there are somc cases that the results of the change of eigenvalues becomes positive or negative by changing the estimator, and there is no reasonable difference in the results between the conventional method, which does not consider the change of neutron source distribution caused by a perturbation, and the new method, which consider that change. Thus it is still necessary to check the Monte Carlo pcrturbation code. Part II. Improvement of Nodal Transport Method for 3-D Hexagonal Geometry It is certain that we can accurately evaluate hexagonal geometry FBR core by nodal transport calculation code for hexagonal-Z geometry named 'NSHEX'. However it is also found that in very heterogeneous core the results is not good enough. Because the treatment of the transverse leakage to the radial direction, which is use for evaluating intra-nodal flux distribution, is not so accurate. For the treatment of the leakage distribution, it is necessary to estimate the nodal vertex flux. In conventional method, the vertex flux estimated by the surrounding node surfacc flux around that vertex. On the contrary,
PNC-TN9410 97-022, 34 Pages, 1997/02
The core averaged burn-up reactivity has been measured and calculated for the Joyo MK-II core. In order to evaluate the relationship between the calculational error of burn-up reactivity and the nuclear data or calculated neutron flux, the burn-up reactivity for an individual fuel subassembly(S/A) must be measured. So the burn-up reactivity measurement test was conducted on the MK-II core. The burn-up reactivity for a driver fuel S/A was measured as a substitution reactivity worth between two S/As at different burn-ups. In the test a fuel S/A with a burn-up of 1 GWd/t was substituted by two S/As with 37 and 62 GWd/t, respectively. The substitutions were carried out at the core center(row 0), middle of the fuel region(row 2)and the border region of the fuel and reflector(row 4). The calculated burn-up reactivity worth performed by the core management code system "MAGI" was compared with the measured value. The results obtained were as follows: (1)Measured substitution reactivity worth(at row 0) between 1 and 37 GWd/t fuel S/A was -0.19%k/kk' and that between 1 and 62 GWd/t was -0.28%k/kk'. (2)Relative distribution of the reactivity worth between 1 and 37 GWd/t agreed with that between 1 and 62 GWd/t. The relative value normalized at core center was 0.67 for the row 2 and 0.28 for the row 4. (3)The C/E value was 1.031.05 for the substitution between 1 and 37 GWd/t and 0.940.95 between 1 and 62 GWd/t at the row 0 and 2. It was clear that the C/E values at the row 4 are higher than those at the row 0 and 2 by 57%. An analysis of the burn-up dependency on the C/E value of the burn-up reactivity worth is being performed in detail. Presents, the PIE of the fuel S/A used for the measurement is under way.