Japan Atomic Energy Research Institute Research Articles

Development of a new integrated IN-VIVO counting system at the QST.

A new in-vivo counting system that functions as both a whole-body counter (WBC) and a lung counter (LC) was developed at the QST to enhance its dose assessment capability. This paper presents an overview of this system and the results of its performance tests. For use of the system as a WBC, three high purity germanium (HPGe) detectors installed in a 20-cm-thick iron shielding chamber are linearly arrayed over a subject lying on the bed, whereas two of the three HPGe detectors are placed over the subject's chest from side to side when using the system as an LC. The new in-vivo system was calibrated using three de-facto phantoms owned by the QST: an adult-male BOttle Manikin ABsorption (BOMAB) phantom, a Lawrence Livermore National Laboratory (LLNL) phantom and a Japan Atomic Energy Research Institute (JAERI) phantom. Monte Carlo simulations were also performed to determine an optimum location for the three detector array in the WBC mode and revealed that the peak efficiency for the BOMAB phantom (662keV) was little varied as long as the middle detector was placed above the thorax and abdomen parts of the phantom. The calculated peak efficiencies agreed well with the observed peak efficiencies for photons with energies over 100keV. For lung counting, a tentative Minimum Detectable Activity of 241Am was evaluated as 9.5Bq for a counting time of 30minutes, and a Japanese male subject with an average chest wall thinness (2.27cm). The developed system is now ready for use.

Validation of thermal–hydraulic system code improved by advanced drift-flux correlation using bundle data at moderate pressure conditions

The accurate prediction of the interfacial drag force is essential to predict the void fraction in a reactor core during the transient and accident phases of nuclear reactors. The key to accurate prediction is to provide reliable drift-flux parameters in calculating the interfacial drag force. The authors previously developed a drift-flux correlation to improve the interfacial drag force prediction accuracy under low-flow and low-pressure conditions. In the present study, the authors demonstrated the validity of the interpolation scheme at a moderate pressure range in the drift-flux correlation and quantified its uncertainty using the experimental data measured in a 5 × 5 rod bundle for the pressures of 0.25, 0.41, 2, 5, and 7 MPa at the multi-purpose Hitachi utility steam test leading facility (HUSTLE). The validated drift-flux correlation was implemented into the RELAP5/MOD3 code and numerical analyses were also performed to reproduce the HUSTLE experimental data. The numerical calculations using the modified RELAP5/MOD3 code showed improved prediction accuracy compared to the calculations by the original RELAP5/MOD3 with the Chexal-Lellouche correlation (e.g., EPRI correlation) in the tested pressure range. The uncertainty of the distribution parameter in the drift-flux correlation was estimated using the HUSTLE database. Next, a numerical analysis was done for the differential pressure in the reactor core in a small break LOCA integral test conducted with the ROSA/large scale test facility (LSTF) by the former Japan Atomic Energy Research Institute. This analysis showed that the predictions with the improved RELAP5/MOD3 agreed with the ROSA/LSTF data within the uncertainty at a 95 % confidence level. It was concluded that the improved RELAP5/MOD3 with the new drift-flux correlation could predict the water level of a reactor core during accident scenarios.

Thermal Hydraulic Analysis and Modeling of the HTTR Using the RELAP5-3D

A thermal hydraulic model for the HTTR (High Temperature Engineering Test Reactor) has been developed using the RELAP5-3D code. The HTTR is a high temperature gas-cooled reactor (HTGR) developed, designed, constructed, and operated by the Japan Atomic Energy Research Institute (JAERI). It is a helium-cooled and graphite-moderated with a thermal power of 30 MW. In this work, a HTTR core inlet and outlet coolant temperatures, mass flow rate, pressure drop and heat structures (fuel and graphite block) temperatures have been reproduced using a RELAP5 model. The simulation was done in the high temperature operation mode (950°C) and in the rated operation mode (850 °C). The results have been compared with the available data. The calculations were performed using the point kinetic model. It is demonstrated that the developed model in the RELAP5-3D is capable of reproduce the thermal behavior of the reactor in steady state operation.

Comparative studies on plutonium utilization in HTTR with helium and carbon dioxide gas coolants

A high temperature engineering test reactor (HTTR) has been developed and constructed by Japan Atomic Energy Research Institute (JAERI) since 1990s. The standard HTTR has been designed for helium coolant with thermal energy output of 30 MW. HTTR uses graphite as the moderator with outlet coolant temperature of 950 °C. Instead of the pebble-bed type of fuel, the prismatic block fuel type with low enriched UO2 has been employed in this reactor.In this study we have performed comparative evaluations on plutonium oxide utilization in uranium fuel cycle of HTTR with helium gas and carbon dioxide gas coolants. The cell-burnup calculations have been performed by using a SRAC 2006 code, with the nuclear data library was derived from JENDL3.3. The 3-dimensional calculations were performed by employing a CITATION code. The results show that the reactor can achieve its criticality condition if the plutonium concentration in the loaded fuel is 5.0% or higher for both coolant materials.

Calibration of Phoswich detector for the measurement of natural uranium in lungs

The occupational workers of fuel fabrication and reprocessing facilities at Indira Gandhi Center for Atomic Research (IGCAR) have a potential for internal exposure to natural uranium which is hazardous both in chemical and radiological aspects. Hence, in vivo monitoring of the radiation workers has to be carried out to ensure safe working conditions. In IGCAR, the in vivo monitoring of natural uranium is being carried out using Phoswich-based lung monitor. The measurement and quantification of internal exposure due to natural uranium is done using 63 and 93 keV photons emitted by 234 Th, immediate daughter of 238 U. Realistic anthropomorphic Lawrence Livermore National Laboratory (LLNL) phantom is used for the calibration of the system. As natural uranium-loaded lung set is not available currently in the laboratory, lung set loaded with indigenously prepared natural uranium source capsules were used. Efficiency curve for 238 U was established for varying muscle equivalent-chest wall thickness (MEQ-CWT) and the efficiency values were in the range of 5.604E-03 to 8.601E-03 CPS/Bq. Simulation results of LLNL voxel phantom having uniform lung distribution of natural uranium agreed within 9% with the measured one which is comparable with the error associated with measurement. This confirms that the distribution pattern of 12 capsules in each lung in the given geometry closely resembles the uniform distribution. The efficiency and minimum detectable activity (MDA) values for Indian population were found to be ranging from 1.011E-03 to 7.931E-03 CPS/Bq and 12–18 Bq, respectively. The efficiency value for 238 U established from the measurement using Japan Atomic Energy Research Institute (JAERI) phantom having uniform source distribution agreed with that of LLNL phantom measurement having capsule source distribution for the same MEQ-CWT thickness (33.8 mm) within 3%. This further reaffirms that the adopted capsule distribution is close to uniform distribution.

Simulation of the ICE P1 test for a validation of COCOSYS and ASTEC codes

Paper presents simulation of the upgraded ICE facility test P1 performed using COCOSYS and ASTEC lumped parameter codes. The experimental ICE facility was constructed at the Japan Atomic Energy Research Institute (JAERI) to obtain experimental data on the ingress-of-coolant event which could occur in the ITER. COCOSYS and ASTEC are computer codes used to simulate accident scenarios in the light water reactors. However, they’ve been also used for fusion safety applications and ASTEC is being extended to better address main accident sequences of fusion installations. It is shown in the paper that pressure and temperature evolutions of the main volumes were predicted with reasonable accuracy, however discrepancies in the suppression and drain volumes suggest underestimated vaporization rate.

Modeling of Zr alloy burst cladding internal oxidation and secondary hydriding under LOCA conditions

The recently developed mechanistic model for Zr alloy cladding hydriding has been implemented in the single-rod SVECHA/QUENCH (S/Q) code. The mass transfer in a fuel rod after ballooning and burst opening have been modeled in the modified code that allowed calculating hydrogen and oxygen pickup by the cladding inner-metal surface. The code predicts with a good accuracy the typical distributions of oxygen and hydrogen in the Zr alloy cladding that were observed in the JAERI (Japan Atomic Energy Research Institute) and ANL (Argonne National Laboratory) single-rod tests and KIT (Karlsruhe Institute of Technology) bundle tests under postulated loss-of-coolant accident (LOCA) conditions.

Some (bacilli) like it hot: genomics of Geobacillus species.

Biotechnology and Biological Sciences Research Council (BBSRC). Grant Numbers: BB/H016120/1, BB/I024631/1, BB/I025956/1, BB/K003240/2, BB/L012499/1

Validation of gadolinium burnout using PWR benchmark specification

The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17×17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor Kinf, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

Development of drift-flux model based on 8 × 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 × 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 × 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data.

JENDL: Nuclear databases for science and technology

It is exactly 50 years since the Japanese Nuclear Data Committee was founded both in the Atomic Energy Society of Japan and in the former Japan Atomic Energy Research Institute. The committee promoted the development of Japan’s own evaluated nuclear data libraries. As a result, we managed to produce a series of Japanese Evaluated Nuclear Data Libraries (JENDLs) to be used in various fields for science and technology. The libraries are categorized into general-purpose and special-purpose ones. The general-purpose libraries have been updated periodically by considering the latest knowledge on experimental and theoretical nuclear physics that was available at the time of the updates. On the other hand, the special-purpose libraries have been issued in order to meet the needs for particular application fields. This paper reviews the research and development for those libraries.

Studies of super-short pulse generated by a free-electron lasers at perfect synchronism

Three-dimensional (3D) simulations and theoretical analyses on super-short pulse generated using freeelectron lasers (FELs) at perfect synchronism are carried out with the help of our 3D OSIFEL code. The evolution of longitudinal pulse width in the Japan Atomic Energy Research Institute (JAERI) experiment is simulated. The results show that the optical pulse is compressed on successive passes due to the slippage between the optical and electron beams, and an ultra-short 221-fs optical pulse is finally obtained, which agrees with the experiment. Furthermore, to shorten wavelength such as soft ultraviolet (SUV) spectrum range, an ultra-short pulse generated at perfect synchronism is analyzed and studied. Finally, the relationship between the optical pulse length compressed and the peak electron beam current is calculated. It shows that the higher the electron beam current, the shorter the output FEL width length, due to the higher gain.

Tumor growth suppression by gadolinium-neutron capture therapy using gadolinium-entrapped liposome as gadolinium delivery agent

Neutron capture therapy (NCT) is a promising non-invasive cancer therapy approach and some recent NCT research has focused on using compounds containing gadolinium as an alternative to currently used boron-10 considering several advantages that gadolinium offers compared to those of boron. In this study, we evaluated gadolinium-entrapped liposome compound as neutron capture therapy agent by in vivo experiment on colon-26 tumor-bearing mice. Gadolinium compound were injected intravenously via tail vein and allowed to accumulate into tumor site. Tumor samples were taken for quantitative analysis by ICP-MS at 2, 12, and 24h after gadolinium compound injection. Highest gadolinium concentration was observed at about 2h after gadolinium compound injection with an average of 40.3μg/g of wet tumor tissue. We performed neutron irradiation at JRR-4 reactor facility of Japan Atomic Energy Research Institute in Tokaimura with average neutron fluence of 2×1012 n/cm2. The experimental results showed that the tumor growth suppression of gadolinium-injected irradiated group was revealed until about four times higher compared to the control group, and no significant weight loss were observed after treatment suggesting low systemic toxicity of this compound. The gadolinium-entrapped liposome will become one of the candidates for Gd delivery system on NCT.

Boron Neutron Capture Therapy for Squamous Cell Carcinoma of the Skull Base

Objective: 1) Describe the first case of inoperative squamous cell carcinoma (SCC) in the skull base recurring after surgery, conventional radiotherapy, and chemotherapy, which was treated using 2-stage boron neutron capture therapy (BNCT). 2) Determine the effectiveness of BNCT for head and neck cancer. Method: We present a case of inoperative recurrent SCC in the right temporal bone involving the occipital condyle. The patient underwent BNCT twice in 1 month in 2007 at Research Reactor No. 4 of the Japan Atomic Energy Research Institute to ensure neutron capture in the deep lesion. Results: An epithermal neutron beam was used as the neutron source. The total estimated radiation dose of the BNCT to the tumor beneath the skin, at the deepest point of the tumor, and in the skin around the right auricle was 41.8, 36.9, and 15.8 Gy-Eq, respectively. After the BNCT, the tumor became necrotic and shrank immediately and disappeared 6 months after the first BNCT. Radiological studies performed 5 years after the first BNCT showed no evidence of residual or recurrent tumors. Conclusion: Total resection of SCC deeply invading the skull base is difficult. Surgical complications can leave patients in critical condition. Two-stage BNCT might be an effective treatment for patients with SCC deep in the skull base.

Design Study of Full Scale Accelerator Driven System (ADS), for Transmuting High Level Waste of MA/Pu

The ADS system used in this study consisting of a high intensity proton linear accelerator, a spallation target, and a sub-critical reactor core. The Pb-Bi spallation target is bombarded by high intensity protons coming from the accelerator. The fast neutrons generated from the spallation reaction were used to drive the sub-critical reactor core. In this ADS system, the neutron source is in the center of reactor core region, so that the neutron distribution was concentrated in the center of core region. In this case, the B/T of MA/Pu could be performed effectively in the center of core region. The neutron energy in the outer region of reactor core was decreased due to the moderation of fuel and coolant materials. Such condition gives a chance to perform Burning and/or Transmutation of LLFPs.The basic parameters of this system are shown in the form of neutronic design, neutron spectrum and B/T rate, including other aspects related to the safety operation system. Furthermore, the analysis of the ADS system was accomplished using ATRAS computer code of the Japan Atomic Energy Research Institute, JAERI[1]. Due to the complexity of the reactor calculation codes, the author has carried out only those calculations needed for analyzing the neutronics system and some parameters related to the safety system. Design study of the transmutation system was a full-scale power level system of 657.53 MWt sub-critical reactor for an accelerator-driven transmutation system. The liquid Pb-Bi was used together as the spallation target materials and coolant of the system, because of some advantages of Pb-Bi in the system concerning the comparison with the sodium coolant. Moreover, they have a possibility to achieve a hard neutron energy spectrum, avoid a positive void reactivity coefficient, allow much lower system operating temperatures, and are favorable for safety in the event of coolant leakage. The multiplication factor of sub-critical core design was adjusted exclusively through the high intensity protons beam accelerator at the spallation target. The fuel was assumed to have homogeneous compositions in the form of (MA-Pu) ZrN mixture with 15 N enriched. The compositions of Pu and MA were the same with the compositions of UO 2 fuel from 33-GWd/t burn-up in PWRs spent fuel after 5 year cooling. The results have been compared with the spent fuel composition from 45 and 60 GWd/t burn-up in PWRs at the same cooling time. The calculation of the burn-up step was 730 days per one batch reloading by using 4-regions core calculation model. The specific parameters of ADS system used in the calculation are described in Table1.

Generation of initial closed flux surfaces by ECH at a conventional aspect ratio of R/a ∼ 3: experiments on the LATE device and JT-60U tokamak

Generation of initial closed flux surfaces by electron cyclotron heating (ECH) at a conventional aspect ratio of R/a ∼ 3 is investigated on the Low Aspect ratio Torus Experiment (LATE) device and further tested on the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) tokamak. In both experiments, a plasma current is initiated and increased by ECH under steady external fields composed of a toroidal field and a weak vertical field. In the LATE experiments, a movable inboard limiter is used to change the aspect ratio of plasmas up to R/a ∼ 3. The results show that the formation of closed flux surfaces is still possible up to R/a ∼ 3 while higher decay indices of vertical field are required as the aspect ratio increases. Similar current start-up discharges are performed on JT-60U, and a plasma current is initiated and increased up to 20 kA under a vertical field having a high decay index.

Re-evaluation of Assay Data of Spent Nuclear Fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries.In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included.This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.

An outcome of nuclear safety research in JAERI: case study for LOCA

An outcome of nuclear safety research (NSR) done by JAERI (Japan Atomic Energy Research Institute) was case studied by the bibliometric method. (1) For LOCA (loss-of-coolant accident) a domestic share of JAERI in monoclinic research paper was 63% at the past (20) 1978-1982 but was decreased to 40% at the present 1998-2002. For co-authored papers a domestic share between JAERI and PS (public sectors) was zero at past (20) but increased to 4% at the present. Research cooperation is active between Tokyo University and JAERI or between JAERI and Nagoya University. (2) It is revealed that LOCA outputs born by NSR-JAERI reflected partly to those of the Safety Licensing Guidelines, however, a share of NSR-JAERI could not determined due to the lack of necessary information in the Guideline.

504 中性子残留応力解析装置RESAの現状(OS5-1 回折法による材料評価)

Diffractometer for residual stress analysis, called RESA, was installed in JRR-3 (Japan Research Reactor No.3) guide hall at JAERI (Japan Atomic Energy Research Institute, now JAEA: Japan Atomic Energy Agency) in 1993 to perform the fundamental studies on the materials engineering. Recently, not only academic users but also industrial users are interested in using the neutron stress measurement technique for ensuring the reliability of mechanical components, and the number of such users is rapidly increasing year by year. To respond to the increase of users of the RESA, the RESA has been upgraded to improve efficiency and versatility of stress measurement, and another neutron diffractometer, RESA-2, has been installed at T2-3-2 beam port in JRR-3 guide hail as well. In this paper, the current status of the RESA-1 and the RESA-2 was introduced.

Trends in cobalt-60 research in nuclear science and technology

This paper quantitatively analyses the growth and development of cobalt-60 research in nuclear science and technology in terms of publications. As per the International Nuclear Information System (INIS) database during 2000-2007, a total of 5874 publications were published. The highest number of papers in a year (896) was in 2002. The average number of publications per year was 734.25. China topped the list with 960 (14.96%) publications followed by Japan with 600 (9.35%), India with 463 (7.21%), Brazil with 339 (5.28%), and the Russian Federation with 266 (4.14%). The authorship and collaboration trend is towards multiauthored papers. Researchers in China, Japan, the Russian Federation and the USA were inclined to work in larger groups. The highly productive institutions were Bhabha Atomic Research Centre, India, with 260 publications; Instituto de Pesquisas Energeticas e Nucleares (IPEN), Brazil with 151 publications; and Japan Atomic Energy Research Institute (JAERI), Japan with 94 publications. The journals most preferred by scientists for the publication of papers were Acta Agriculturae Nucleatae Sinica with 183 publications, followed by Chinese Journal of Radiological Medicine and Protection with 182 publications and Radiation Physics and Chemistry with 175 publications.