Development of calculation tool of neutron flux used for evaluation of radioactivity inventory in radioactive waste generated from research, medical and industrial facilities

Kurosawa, Ryohei; Okada, Shota; Sakai, Akihiro; Nakata, Hisakazu; Amazawa, Hiroya

JAEA-Data/Code 2015-005, 82 Pages, 2015/06

JAEA-Data-Code-2015-005.pdf:3.47MB

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.

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC TN9400 2000-037, 87 Pages, 2000/03

JNC-TN9400-2000-037.pdf:3.48MB

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 ...

Analyses of neutron and ray measurements in a target room of several tens MeV Proton Facility

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

https://doi.org/10.1080/00223131.2000.10874873

no abstracts in English

Preliminary experiment of neutron capture cross section of Tc-99 with lead slowing-down spectrometer

*

JNC TJ9400 2000-009, 63 Pages, 2000/02

JNC-TJ9400-2000-009.pdf:2.48MB

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.

Study on dose distributions in a slab phantom irradiated by several tens MeV neutrons

Nakane, Yoshihiro

JAERI-Research 2000-003, p.110 - 0, 2000/02

JAERI-Research-2000-003.pdf:5.55MB

no abstracts in English

Preparation of methods to calculate pin-wise intra-subassembly power density distribution of a new in-pile experimental reactor for FBR safety research

Mizuno, Masahiro*; Uto, Nariaki

JNC TN9400 98-007, 147 Pages, 1998/11

JNC-TN9400-98-007.pdf:8.32MB

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 ...

Experimental analysis of coherent neutron flux fluctuations observed in a pressurized water reactor

Suzudo, Tomoaki; Tuerkcan, E.*; H.Verhoef*

Nuclear Science and Engineering, 129(2), p.203 - 208, 1998/06

https://doi.org/10.13182/NSE98-A1976

no abstracts in English

Super-Phenix Benchmark used for Comparison of PNC and CEA Calculation Methods,and of JENDL-3.2 and CARNAVAL IV Nuclear Data

Hunter

PNC TN9410 98-015, 81 Pages, 1998/02

PNC-TN9410-98-015.pdf:3.15MB

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 ...

Evaluation of nuclear characteristics of DCA modification core for sub-critica1 measurement

Hazama, Taira

PNC TN9410 97-088, 139 Pages, 1997/10

PNC-TN9410-97-088.pdf:3.01MB

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.

Improvement of numerical analysis method for FBR core characteristics (II)

Takeda, Toshikazu*; *; Kitada, Takanori*; *

PNC TJ9605 97-001, 100 Pages, 1997/03

PNC-TJ9605-97-001.pdf:2.82MB

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,