Kohtaro Ueki

Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

Abstract The Monte Carlo coupling technique with the coordinate transformation is used to evalulate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the “SSW-SSR/CRT calculation system.” In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements. It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house.

Analyses and Development of Effective Compensation Shields in a Multilegged Duct Streaming System

Abstract The 14-MeV neutron streaming experiment with three straight ducts is analyzed with the cell flagging technique of the MCNP code. The contributions of neutrons passing through the flagging cell located at the duct inlet and entering the detectors located in the duct are identified quantitatively. Furthermore, the streaming paths of neutrons entering in the duct inlet are cleared by the analysis. As an application of the cell flagging technique, a neutron streaming system with two-bend cylindrical duct in a thick concrete shield is prepared, and the flagging cells are located around the duct. The contributions of neutrons passing the flagging cells to the detector located at the duct outlet are cleared, and effective compensation shields to reduce the neutron dose-equivalent rate at the duct outlet are obtained by replacing some of the flagging cells with the NS-4-FR shield. Moreover, it is expected that the equilibrating contribution from each flagging cell to the dose-equivalent rate at the duct outlet is the essential function to make an effective compensation shielding system with neutron streaming.

Integral Test of JENDL-3.3 with Shielding Benchmarks

Integral test of neutron and gamma-ray production data for the latest version of Japanese Evaluated Nuclear Data Library, Version 3.3 (JENDL-3.3) has been performed by using shielding benchmarks. An evaluation scheme for shielding benchmark analysis established in Japanese Nuclear Data Committee (JNDC) was applied to the integral test for medium-heavy nuclei such as Aluminum, Sodium, Titanium, Vanadium, Chromium, Iron, Cobalt, Nickel, Copper, Niobium and Tungsten. Calculations were made based on a continuous-energy Monte Carlo code MCNP4B/4C and multi-group discrete ordinates codes ANISN and DORT. The latest version of NJOY99 was employed to generate cross-section libraries for these transport codes. Calculations with JENDL-3.2, ENDF/B-VI, FENDL-1 and FENDL-2 were also made for comparison. In the present study, benchmark results of neutron and gamma-ray production data are shown for Sodium, Iron, Vanadium, Tungsten, Nickel, Titanium, Chromium, Niobium and Cobalt

Optimum Arrangement to Minimize Total Dose Rate of Iron-Polyethylene Shielding System

The shielding experiments using iron-polyethylene slab shields with a 252Cf neutron source were carried out to find out an optimum arrangement to minimize the total dose rate; that is composed of neutron and secondary γ-ray dose rates. The total thickness of the iron slabs was fixed at 32 cm, while a variety of thickness and location of a polyethylene slab in the iron slabs were employed as a parameter. The minimum dose point (i.e. optimum shielding arrangement) was observed when the polyethylene slab was located at approximately 20 cm depth from the source side in the arrangements. The ratio of the minimum dose rate obtained for the optimum arrangement to the maximum for the worst arrangement became 1/1.3 for the polyethylene slab of 1-cm-thick, 1/2.0 for 3-cm-thick, 1/2.9 for 6-cm-thick, 1/3.9 for 10-cm-thick and 1/3.6 for 14-cm-thick. The appearance of the optimum arrangement for the total dose rates, and changing profiles of the secondary γ-ray as well as the neutron dose rates were reproduced for the...

Combination of four biasing techniques for gamma ray shielding calculations

To enhance the efficiency the of Monte Carlo method for particles in deep penetration problems for complex geometry, we have used a method based on a combination of four techniques: the exponential transformation, the angular biasing, the imposed collision and the weight window. These four techniques allowed us to increase the probability of particles to scatter toward a finite detector. To test the effectiveness of this method we have applied it to a practical problem. We are interested in the evaluation of the gamma ray which can skirt a lead (Pb) shield within a graphite medium and contribute to a finite detector, placed behind this presupposed perfect shield. A punctual, isotropic and mono-energetic gamma source is placed at the other side of the shield. In order to qualify this method we have used the TRIPOLI-03 and MCNP-4B codes. The current obtained from our multi-group Monte Carlo program agrees with both codes for different locations of the detector with high Figure of Merit (FOM = 1σ2T). The used gamma ray cross-sections were collapsed to 75 groups from the ENDF/B-VI library. This method has also been used successfully to calculate the gamma ray spectra for different locations of the detector.

Validation of gamma-ray production data of iron in JENDL-3.2 with shielding benchmark

An integral test of γ-ray production data of iron in the latest version of Japanese Evaluated Nuclear Data Library (JENDL-3.2) has been performed by means of a shielding benchmark analysis of KfK leakage neutron and γ-ray spectrum measurements from iron spheres with a 252Cf source in the center. Two comprehensive systems which consist of a continuous-energy Monte Carlo method and a multi-group Sn transport method have been adopted in this benchmark analysis. For comparison, analyses with JENDL-3.1, FENDL-1 and ENDF/B-IV have been also carried out. The calculation using JENDL-3.2 showed a good agreement with the experiment. It has been concluded that the γ-ray production data of iron in JENDL-3.2 were applicable for use of shielding designs and analyses of the fission neutron source problem.

Overview of Recent Research Activities of Monte Carlo Simulation in Japan

This paper describes recent progresses of the Monte Carlo simulation technology in nuclear energy field in Japan. Radiation shielding solution method using the Monte Carlo had been validated as a reliable tool through the discussion of “Radiation Shielding Safety Demonstration Analysis Group” of Japan Atomic Energy Research Institute. Since 1996, “Monte Carlo Simulation Working Group” has been accumulating use experiences of Monte Carlo codes in the wide range of nuclear energy field. This working group is planing to publish “Guideline of Monte Carlo Simulations” during FY-99. This “Guideline” is expected to be a first Japanese practical textbook of Monte Carlo calculation. In 1998, the first full-scale topical conference on Monte Carlo simulation was held in Tokyo. “Research Committee on Particle Simulation with the Monte Carlo Method” was established in Atomic Energy Society of Japan in 1998. This committee is composed of more than seventy members from many fields of nuclear energy research in Japan. This committee is expected to be a core that will drive the research and development activity of Monte Carlo calculation in Japan.

Biasing Techniques for Gamma Rays going around Efficient Shields

This paper describes a method based on a combination of the exponential transformation, the angular biasing and the region of imposed collision. This combination can be employed in Multigroup Monte Carlo radiation transport calculations particularly in deep penetration problems for complex geometry. To test the effectiveness of this method, we have applied it to a practical case concerning the evaluation of gamma rays, which skirt a region of perfect shield within a graphite medium and contribute to a finite detector, placed behind the perfect shield. An isotropic punctual and mono-energetic gamma source is placed at the other side of the shield. The current obtained for our multigroup Monte Carlo program agrees with MCNP4B code with a high figure of Merit. The gamma ray cross section used was collapsed to 75 groups from ENDF/B-VI library.

Measurement of Dose-Equivalent Rates around a Cask and Monte Carlo Analysis with Actual Configuration of Fuel Basket

Gamma ray and neutron dose-equivalent rate distributions are measured with an ionization-type gamma-ray survey meter and a moderator-type neutron survey meter respectively around a TN-12A spent fuel transport cask, and the measured dose-equivalent rate distributions are analyzed by the Monte Carlo method. Two models for the aluminum-alloy fuel basket are considered in the Monte Carlo code MCNP 4B. In one case the configuration of the basket with 12 spent fuel assemblies is modeled in detail and the other is the homogenized basket as employed in the Sn code DOT 3.5. In addition, the burn-up distribution is taken into account to generate source neutrons and gamma rays in the z-axis of the spent fuel assemblies in both cases. The essential difference in the dose-equivalent rates is obtained from the Monte Carlo calculations employing the homogenized model and the actual configuration of the basket. Due to employing the actual configuration, the gamma ray and neutron dose-equivalent rates reduce to 67% and 80%, respectively as compared with the homogenized model on the surface of the TN-12A cask. As the results, the good C/Es are obtained: for the neutron it is approximately 1.0 and 1.25 for the gamma ray it is at the center of the cask surface, respectively The effect of the burn-up distribution appears clearly at the off-center cask surface, and in particular, the neutron dose-equivalent rates come close to the measured ones.

Measurements and Calculations of the Dose Distribution from Co-60 Gamma Rays in Multiple Straight Ducts through Iron Shields

Dose distribution in multiple straight ducts in iron shields was measured for gamma rays from a 60Co source using LiF:Mg, Cu, P thermoluminescent dosimeters. The basic iron shield consisted of a 50×50×25 cm3 iron block through which 100 straight ducts separated by 4 cm were drilled. Two shields were prepared, one with duct diameter of 0.95 cm and the second with duct diameter of 2.0 cm. The measured doses agreed with calculations made using the MCNP Monte Carlo transport code. The radiation in the ducts were separated into three components: a streaming component coming through the duct in question, a diffusion component consisting of radiation coming through the bulk iron shield and a “flowing” component coming from radiation streaming through adjacent ducts. From calculations based on bulk shields, single ducts and an 8-duct group surrounding the duct in question, it was possible to estimate the effect of adjacent ducts as approximately 33% for the 0.95 cm diameter ducts and 61% for 2.0 cm diameter ducts at maximum. By flagging photons passing through the mouths of the cells in the MCNP calculations for the 2.0 cm ducts, it was noticed that the contribution from unscattered photons was large. Calculations were also made with the QAD point-kernel code: these gave doses which were the same or slightly larger than those measured, except at depths of 25 cm.