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Journal Articles

Medical application of Particle and Heavy Ion Transport code System PHITS

Furuta, Takuya; Sato, Tatsuhiko

Radiological Physics and Technology, 14(3), p.215 - 225, 2021/09

Number of the PHITS users has steadily increased since 2010 from when it is officially counted. Among them, increase of new users in medical physics is outstanding. Many research works in medical physics using PHITS have been published and the applications are widely spread in different fields such as applications to different types of radiotherapy, shielding calculations of medical facilities, application to radiation biology, and research and development of medical tools. In this article, we will introduce useful functions for medical application in PHITS by referring to examples of various medical applications.

Journal Articles

A New convention for the epithermal neutron spectrum for improving accuracy of resonance integrals

Harada, Hideo; Takayama, Naoki; Komeda, Masao

Journal of Physics Communications (Internet), 4(8), p.085004_1 - 085004_17, 2020/08

A new convention of epithermal neutron spectrum is formulated for improving accuracy of resonance integrals. The new type function is proposed as an approximating function of epithermal neutron spectrum based on calculations by the state-of-art Monte Carlo code MVP-3. Bias effects on determination of resonance integrals due to utilizing approximating functions of the traditional types and the new type are compared. The other bias effect is also investigated, which is caused by neglecting position dependence of a neutron spectrum inside an irradiation capsule. For demonstrating the bias effects due to these assumptions on neutron spectrum quantitatively in a practical case, the thermal neutron-capture cross section and resonance integral of $$^{135}$$Cs measured at a research reactor JRR-3 are re-evaluated. A superior property of the proposed new convention is discussed. The experimental method is proposed to determine the new shape factor $$beta$$ introduced in the convention by a combinational use of triple flux monitors ($$^{197}$$Au, $$^{59}$$Co and $$^{94}$$Zr), and its analytical methodology is formulated.

Journal Articles

Gamma detector response simulation inside the pedestal of Fukushima Daiichi Nuclear Power Station

Riyana, E. S.; Okumura, Keisuke; Terashima, Kenichi; Matsumura, Taichi; Sakamoto, Masahiro

Mechanical Engineering Journal (Internet), 7(3), p.19-00543_1 - 19-00543_8, 2020/06

JAEA Reports

Study on control rod model in HTTR core analysis

Nagasumi, Satoru; Matsunaka, Kazuaki*; Fujimoto, Nozomu*; Ishii, Toshiaki; Ishitsuka, Etsuo

JAEA-Technology 2020-003, 13 Pages, 2020/05

JAEA-Technology-2020-003.pdf:1.5MB

The influence of the control rod model on the nuclear characteristics of the HTTR has been evaluated, by creating detailed control rod model, in which geometric shape was close to that of the actual control rod structure, in MVP code. According to refinement of the control rod model, the critical control rod position was 11 mm lower than that of the conventional model, and this was close to the measured value of 1775 mm. The reactivity absorbed by the shock absorber located at the tip of the control rod was 0.2%$$Delta$$k/k, and this was 14 mm difference at the critical control rod position. Considering the effect of refinement of the control rod and the effect of the shock absorber, the correction amount for the analysis value in SRAC code due to the shape effect of the control rod, is -0.05%$$Delta$$k/k in reactivity, and -3 mm in the critical control rod position at low temperature criticality.

Journal Articles

Calculation of gamma and neutron emission characteristics emitted from fuel debris as a basis for determination of suitable detector system

Riyana, E. S.; Okumura, Keisuke; Terashima, Kenichi

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05

Journal Articles

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

JAEA Reports

MVP/GMVP version 3; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

JAEA-Data/Code 2016-018, 421 Pages, 2017/03

JAEA-Data-Code-2016-018.pdf:3.89MB
JAEA-Data-Code-2016-018-appendix(CD-ROM).zip:4.02MB
JAEA-Data-Code-2016-018-hyperlink.zip:1.94MB

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.

Journal Articles

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

 Times Cited Count:1 Percentile:14.27(Nuclear Science & Technology)

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.

JAEA Reports

SWAT4.0; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*

JAEA-Data/Code 2014-028, 152 Pages, 2015/03

JAEA-Data-Code-2014-028.pdf:13.39MB

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0.

JAEA Reports

MVP/GMVP 2; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

JAERI 1348, 388 Pages, 2005/06

JAERI-1348.pdf:2.02MB

To realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector supercomputers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them.

Journal Articles

Impact of perturbed fission source on the effective multiplication factor in Monte Carlo perturbation calculations

Nagaya, Yasunobu; Mori, Takamasa

Journal of Nuclear Science and Technology, 42(5), p.428 - 441, 2005/05

 Times Cited Count:50 Percentile:95.28(Nuclear Science & Technology)

A new method to estimate a change in the effective multiplication factor due to the perturbed fission source distribution has been proposed for Monte Carlo perturbation calculations with the correlated sampling and differential operator sampling techniques. The method has been implemented into the MVP code for verification. Simple benchmark problems have been set up for fast and thermal systems and the applicability of the method has been verified with the problems. In consequence, it has been confirmed that the method is very effective to estimate the change. It has been also shown that there are some cases where the perturbed source effect is significant and the change in reactivity cannot be estimated accurately without taking the effect into account. Even in such cases, the new method can estimate the perturbed source effect and the estimation of the change in reactivity has been remarkably improved.

Journal Articles

Examination for neutron dose assessment method from induced sodium-24 in human body in criticality accidents

Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro

Journal of Nuclear Science and Technology, 42(4), p.378 - 383, 2005/04

 Times Cited Count:2 Percentile:18.81(Nuclear Science & Technology)

Experiments were made to verify a dose assessment method from activated sodium in body in criticality accidents. A phantom containing sodium chloride solution was irradiated in the Transient Experiment Critical Facility to simulate activation of sodium. Monte Carlo calculations were performed to obtain quantitative relation between the activity of induced Na-24 and neutron dose in the phantom. In the previous work, conversion coefficients from specific activity of induced Na-24 to neutron dose had been analyzed with the MCNP-4B code concerning neutron spectra at some hypothesized configurations. One of the prepared coefficients was applied to evaluate neutron dose from the measured activity. The estimated dose agreed with the dose analyzed by the Monte Carlo calculation in the present study within an acceptable uncertainty, which is indicated by the IAEA. In addition, the dose calculated with the prepared coefficient was close to the result measured with dosimeters. These results suggest that the prepared coefficients can be applied to dose assessments from induced Na-24 in body.

Journal Articles

An Improved fast neutron radiography quantitative measurement method

Matsubayashi, Masahito; Hibiki, Takashi*; Mishima, Kaichiro*; Yoshii, Koji*; Okamoto, Koji*

Nuclear Instruments and Methods in Physics Research A, 533(3), p.481 - 490, 2004/11

 Times Cited Count:3 Percentile:25.71(Instruments & Instrumentation)

The validity of a fast neutron radiography quantification method, the $$Sigma$$-scaling method, which was originally proposed for thermal neutron radiography was examined with Monte Carlo calculations and experiments conducted at the YAYOI fast neutron source reactor. Water and copper were selected as comparative samples for a thermal neutron radiography case and a dense object, respectively. Although different characteristics on effective macroscopic cross-sections were implied by the simulation, the $$Sigma$$-scaled experimental results with the fission neutron spectrum cross-sections were well fitted to the measurements for both the water and copper samples. This indicates that the $$Sigma$$-scaling method could be successfully adopted for quantitative measurements in fast neutron radiography.

Journal Articles

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.

Journal Articles

A Study on induced activity in the low-activationized concrete for J-PARC

Matsuda, Norihiro; Nakashima, Hiroshi; Kasugai, Yoshimi; Sasamoto, Nobuo*; Kinno, Masaharu*; Kitami, Takayuki; Ichimura, Takahito; Hori, Junichi*; Ochiai, Kentaro; Nishitani, Takeo

Journal of Nuclear Science and Technology, 41(Suppl.4), p.74 - 77, 2004/03

In high power proton accelerator facilities, concrete shield can be highly activated, which makes maintenance work quite difficult. So, a low-activationized concrete (limestone concrete) is to be partially adopted as a concrete shield for Japan Proton Accelerator Research Complex (J-PARC) aiming at reducing $$gamma$$-ray exposure dose during maintenance period. A new quantity, $$^{24}$$Na-equivalent, was introduced as a criterion to assure effectiveness of the low-activationized concrete. In order of its verification, powdered low-activationized concrete and ordinary one were irradiated using FNS at JAERI. The measurements were analyzed by a shielding design code system being used for J-PARC, showing that the calculations reproduce the measured induced activity within a factor of 2. Furthermore, by using the same code system, $$gamma$$-ray exposure dose was calculated for the configuration of J-PARC to find out that $$gamma$$-ray exposure dose by the low-activationized concrete was about 10 times lower than that by the ordinary concrete in a period of less than a few days after operation.

Journal Articles

3-D shielding calculation method for 1-MW JSNS

Maekawa, Fujio; Tamura, Masaya

Proceedings of ICANS-XVI, Volume 3, p.1051 - 1058, 2003/07

A three-dimensional (3-D) shielding calculation model for MCNPX was produced for shielding design of 1-MW JSNS. The model included simplified target-moderator-reflector assembly, helium-vessel and neutron beam extraction pipes, shutters, shield blocks, gaps and void spaces between these components, and so on, and could treat streaming effects precisely. The particle splitting and kill method with cell importance parameters was adopted as a variance reduction method. The cell importance parameters for such a large target station of about 15 m in diameter and 12 m in hight in which neutron fluxes attenuated more than 12 orders of magnitude could be determined appropriately by automated iteration calculations. This calculation procedure enabled detailed 3-D shielding design calculations for the whole target station in a short time, i.e., within 2 days, and contributed for progress of shielding designs of JSNS.

Journal Articles

Effect of phantom material on backscattered radiation against photon irradiation

Takahashi, Fumiaki; Yamaguchi, Yasuhiro

Radioisotopes, 52(2), p.94 - 97, 2003/02

Effect of phantom material on backscattered radiations was studied for photon irradiation. Monte Carlo calculations using MCNP-4B code were performed to analyze scattered radiation on the surface of 30x30x15cm3 slab phantoms with different material. Dose on the surface of a human body was also estimated with a modified MIRD-5 type phantom. No significant difference of dose due to scattered radiation was observed between a soft tissue slab and phantom the water-filled slab phantom recommended by the International Organization for Standardization. On the other hand, dose on the surface of the PMMA phantom was found to be larger than doses on the phantom with water or soft tissue. The results also showed that response of dosimeter on the ISO phantom would be near to that on the trunk of a human body.

Journal Articles

Comparison of synchrotron radiation calculation between EGS4, FLUKA, PHOTON and STAC8

Asano, Yoshihiro; Liu, J. C.*

KEK Proceedings 2002-18, p.48 - 54, 2003/01

The light energy was not higher than soft X-rays so that the radiation safety for synchrotron radiation passed for unworthy issues of attention. With constructing and operating the third generation facilities such as SPring-8, however, high energy synchrotron radiation of extremely high intense is available, and high accuracy should be required in the shielding and safety design for the beam of synchrotron radiation. Furthermore, the middle size facilities such as CLS (Canadian Light Source) are now under construction and the optimum design of shielding is desired to construct the beamline with safety and compact. It is necessary for the purpose to verify the detailed evaluation technique. Therefore, taking the SPEAR3 bending magnet and BL11-3 wiggler beamlines in the examples, the characteristics of the codes were investigated in comparison with synchrotron radiation shielding design code PHOTON, STAC8 and Monte Carlo simulation code EGS4, FLUKA

Journal Articles

Analyses of absorbed dose to tooth enamel against external photon exposure

Takahashi, Fumiaki; Yamaguchi, Yasuhiro; Iwasaki, Midori*; Miyazawa, Chuzo*; Hamada, Tatsuji*; Funabiki, Jun*; Saito, Kimiaki

Radiation Protection Dosimetry, 103(2), p.125 - 130, 2003/01

 Times Cited Count:4 Percentile:32.68(Environmental Sciences)

Absorbed dose to tooth enamels against external photon exposure was examined by the Electron Spin Resonance (ESR) dosimetry using tooth samples placed in a realistic physical phantom. Dose to teeth region was also measured with thermo-luminescence dosimeters (TLDs). A voxel-type phantom was constructed from CT images of the physical phantom. Monte Carlo calculations with this voxel-type phantom were performed to analyse the results of the experiments. The obtained data in this study were compared to the enamel doses, which were calculated with a modified MIRD-type and already given in a previous paper. The results suggested that the conversion factors from enamel dose to organ doses obtained by the modified MIRD-type phantom are to be applicable for retrospective individual dose assessments by the ESR dosimetry. The analysis, however, indicated that the size and figure of the head can affect the enamel dose for low photon energy region below 100keV.

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