Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 29

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Development of human resource capacity building assistance for nuclear security

Noro, Naoko; Nakamura, Yo; Hirai, Mizuki; Kobayashi, Naoki; Kawata, Norio; Naoi, Yosuke

Modern Environmental Science and Engineering, 3(5), p.309 - 313, 2017/05

The paper introduces methodologies of ISCN for nuclear security training curricula development. The paper focuses on the contribution of ISCN for curriculum development of IAEA transport security training course which ISCN hosted in 2015.

Journal Articles

Development of human resource capacity building assistance for nuclear security

Noro, Naoko; Nakamura, Yo; Hirai, Mizuki; Kobayashi, Naoki; Kawata, Norio; Naoi, Yosuke

Proceedings of 18th International Symposium on the Packaging and Transport of Radioactive Materials (PATRAM 2016) (DVD-ROM), 7 Pages, 2016/09

The paper describes the methdologies ISCN has adopted to develop its training curricula on nuclear security. ISCN has been working closely with the international partners to build its own capacity, especially with the Sandia National Laboratories of the United States. The paper then focuses on the joint course development between IAEA and ISCN on the International Training Course on Transport Security which ISCN hosted in 2015.

Journal Articles

Nuclear security culture in comparison with nuclear safety culture; Resemblances and differences

Kawata, Norio

Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-36-Kai Nenji Taikai Rombunshu (Internet), 7 Pages, 2015/12

Since the terrorist attacks on the U.S. on September 11th,2001, Nuclear Security has been focused on and treated as a global issue in the international community and it has also been discussed as a real and serious threat to nuclear power plants in the world since The Great East Japan Earthquake in March, 2011. The International Atomic Energy Agency (IAEA) issued a document including Nuclear Security Recommendations (INFCIRC/225/Rev.5) (NSS 13) in the Nuclear Security Series and emphasized the necessity of fostering Nuclear Security Culture. Nuclear Security Culture has been frequently discussed at various kinds of seminars and events. Since the officials in charge of Nuclear Security are familiar with the area of Nuclear Safety, the relationships between Nuclear Safety Culture and Nuclear Security Culture have been the point in controversy. This paper clarifies relevance between Nuclear Safety and Nuclear Security, considers resemblances and differences of their concepts and lessons learned for each culture from nuclear power plant accidents, and promotes deeper understanding of Nuclear Safety and Nuclear Security Culture.

Journal Articles

Evaluating methods to improve safeguards training courses of ISCN

Okumura, Yukiko; Nakamura, Yo; Kawata, Norio

Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-35-Kai Nenji Taikai Rombunshu (Internet), 8 Pages, 2015/01

Although questionnaires were used to receive feedbacks from participants at the end of each training course, Integrated Support Center for Nuclear Nonproliferation and Nuclear Security (ISCN) of Japan Atomic Energy Agency (JAEA) did not establish a structured evaluation method. To this end, ISCN has started to study on methods to accurately evaluate the courses since April and started to introduce the evaluation method on trial, according to the Donald Kirkpatrick's Four-Level Training Evaluation Model, so as to better develop and conduct more effective courses. This paper will focus on how ISCN has modified the Kirkpatrick's Four-level to adapt to its safeguards training courses. This will then be followed by two particular cases of how the evaluation method functioned for the Additional Protocol training courses held in Malaysia in 2014, and the feedbacks received to improve future training courses.

Journal Articles

Development of the training tools for nuclear security; Physical Protection Exercise Field (PPEF) and Virtual Reality (VR) training system

Kawata, Norio; Wakabayashi, Shuji; Hanai, Tasuku; Yamaguchi, Yasuo; Nonaka, Nobuyuki; Scharmer, C.*

Proceedings of International Conference on Nuclear Security; Enhancing Global Efforts (CD-ROM), 10 Pages, 2014/03

The ISCN of the JAEA provides effective trainings in order to strengthen nuclear security for emerging nuclear power countries in Asia to realize Japan's National Statement at the Washington Nuclear Security Summit in April 2010. As a part of these activities, the ISCN has developed the PPEF and the VR training system, which are training tools to implement experience-oriented and interactive lessons. These two facilities are mutually complemented and contribute to deeper understandings through actual practices in addition to the classroom lesson. The ISCN initiated its full-scale training from 2012 JFY, and these two facilities received more than 450 trainees or vistors from Japan and over-sea countries. This paper provides the basic concepts and outlines of these two facilities and the training programs that use them for teaching the nuclear security.

Journal Articles

Outline of the SSAC training course

Iwai, Naofumi; Kuribayashi, Toshihiro; Kawata, Norio

Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-34-Kai Nenji Taikai Rombunshu (Internet), 4 Pages, 2013/10

In order to support international safeguards activities conducted by the IAEA, the Integrated Support Center for Nuclear Nonproliferation and Nuclear Security (ISCN) of the Japan Atomic Energy Agency (JAEA) has been providing the International Training Course on State Systems of Accounting for and Control of Nuclear Material (SSAC) targeted mainly at Asian nations as a Japan's subsidized project by making good use of Japan's efforts and experience in nonproliferation. The course provides the opportunity of a facility tour and an A-bombed site visit, as well as offers lectures and practical trainings necessary to build and maintain the SSAC for those including "government officials involved in safeguard policy and regulation" and "business operators engaged in accounting". This paper discusses the summary of the SSAC training course operation held at the ISCN.

Journal Articles

Outline of physical protection exercise field

Kawata, Norio; Wakabayashi, Shuji; Naito, Aisaku

Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-33-Kai Nenji Taikai Rombunshu (Internet), 6 Pages, 2012/10

The Integrated Support Center for Nuclear Nonproliferation and Nuclear Security (ISCN) of the Japan Atomic Energy Agency set up exercise facilities for trainee of nuclear power emerging countries in Asia involved in Physical Protection (PP) including government officers in charge of nuclear security policy or nuclear security regulation, planning and management staff of PP facilities of operating companies, design professionals for PP facilities, and security personnel responsible for PP. After April in 2012, the facility started to be applied to actual ISCN's PP training and is expected as training field for not only Asian nuclear emerging country but also domestic nuclear energy companies and regulatory bodies. In order to provide effective and practical exercises, we set up the training facilities with basic measures and equipment typical of those used in actual PP facilities, e.g., protective fences, sensors, and cameras. This paper provides an outline of the facilities.

Journal Articles

The Current status of the Russian surplus weapons plutonium disposition and its Japanese contribution

Funada, Toshio; Kawata, Norio; Senzaki, Masao

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 50(11), p.711 - 715, 2008/11

no abstracts in English

Journal Articles

Recent status and prospects of Russian surplus weapons plutonium disposition

Kawata, Norio; Yano, Soichiro

Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-27-Kai Nenji Taikai Rombunshu (CD-ROM), 9 Pages, 2006/00

Russian Surplus W-Pu was produced by byproducts of disarmament of nuclear weapon and recognized as a serious threat of nuclear proliferation. US and RF concluded an agreement to dispose of no less than 34 ton-Pu each in 2000. G7 countries including Japan jointed to collaborative investigation to dispose of Russian W-Pu. Relative countries have been discussing the scenario of Pu disposition mainly by burning in LWR (VVER1000) with MOX pellet fuel and FR (BN600) with MOX vipac fuel, however actual disposition work faced dead lock since estimated necessary cost is more than 3 times of pledged finance by G7 countries. JAEA has conducted to realize BN600 vipac fuel option, which is MOX vipac fuel made with W-Pu is burned in BN600 reactor, based on experiences of Pu utilization and fast reactor technology in a decade years. Technical feasibility is established through several cooperative studies with Russian Institutes. Since this option is confirmed as a low-cost and early disposition method, recently, US and RF seek to accelerate this option prior to VVER1000 option. This paper will provide a description of progress on BN600 vipac fuel option and future prospects of whole plan of Russian Surplus W-Pu disposition of 34 ton.

Journal Articles

Study of In-Pile Test Facility for Fast Reactor Safety Research: Performance Requirements and Design Features

; ; ; ;

Proceedings of ENS Class 1 Topical Meeting; Research Facilities for the Future of Nuclear Energy, p.512 - 52, 1996/00

None

Journal Articles

Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

Uto, Nariaki; Ohno, Shuji;

Proceedings of International Conference on the Physics of Reactors (PHYSOR '96), 0 Pages, 1996/00

None

JAEA Reports

None

Niwa, Hajime; Kawata, Norio; Ieda, Yoshiaki; Sato, Ikken; Ohno, Shuji; Uto, Nariaki; Miyahara, Shinya; Kondo, Satoru; Kamide, Hideki; Yamaguchi, Akira; et al.

PNC TN9410 94-154, 317 Pages, 1995/03

PNC-TN9410-94-154.pdf:13.66MB

None

JAEA Reports

The Calculation accuracy analysis of the core design code for ATR (Advanced thermal reactor demonstration plant; Reactivity worths of SUS control rod, Boron in D$$_{2}$$O and Xenon

*; *

PNC TN9410 89-114, 140 Pages, 1989/08

PNC-TN9410-89-114.pdf:4.99MB

The calculation accuracy evaluation of the SUS control rod reactivity worth, boron reactivity and Xenon reactivity were performed on the basis of DCA experimental data, "Fugen" start-up test data and "Fugen" operation data by using the WIMS-ATR code. The main results obtained are as follows. (1)As is shown in the results of the calculation accuracy evaluation of the channel power distribution for DCA Pu-U two-region core by the WIMS-ATR code, the RMS(Root Mean Square) of the calculation error of the channel power distribution is 4.8% and the relative error of the radial peaking factor is +2.1%. (2)As is shown in the results of the calculation accuracy evaluation of the SUS control rods worths by the WIMS-ATR code, the calculation error of the SUS control rod worth is -5,9%, and the value satisfy the design margin +2.0,-1.9 % (C/E-1). (3)As is shown in the results of the calculation accuracy evaluation of the boron (in the heavy water) reactivity based on "Fugen" start-up test data by the WIMS-ATR code / the LAYMON-2A code, the calculation error of the boron reactivity is-2.2%$$sim$$+7.9%, and the value satisfies the design margin of $$pm$$10%. (4)As is shown in the results of the calculation accuracy evaluation of the Xenon reactivity based on "Fugen" operation data by the WIMS-ATR code/the LAYMON-2A code, the calculation error of the Xenon reactivity is-1.20%$$sim$$+14.7%, and the value satisfies the design margin $$pm$$20%.

JAEA Reports

The Calculation accuracy analysis of the core design code for ATR (Advanced Thermal Reactor) demonstration plant; Control rod reactivity worth and DCA power distribution

*; *; *

PNC TN9410 88-109, 171 Pages, 1988/08

PNC-TN9410-88-109.pdf:9.91MB

The calculation accuracy evaluation of the control rods reactivity worths and channel power distributions were performed on the basis of DCA experimental data and "Fugen" start-up test data by using the WIMS-ATR code. The main results which have been obtained are as follows. (1)It is proved that calculation values by the WIMS-ATR code over-estimates the B$$_{4}$$C control rods worths about 4%-9% compared with the METHUSELAH-II code. (2)The calculation accuracy of the B$$_{4}$$C control rods worths based on the "new" absorption area method used cell averaged diffusion constants by the WIMS-ATR code is about -6% $$sim$$ +8%. From this result, it is proposed to use $$pm$$10% as the design margin to the control rod reactivity worths. (3)It is considered that the multi-cell method using imhomogenious nuclear constants by using the WIMS-ATR code is able to estimate the control rods reactivity worths with the errors of about $$pm$$5%. Therefore, it is proposed to use the multi-cell analysis method as a detailed calculational method for reevaluation of design values for control rods reactivity worths. (4)The RMS (Root Mean Square) of the calculation errors of the cannel power distributions is about 2 $$sim$$ 3%. Therefore, it is shown that the calculational values by the WIMS-ATR code are in good agreement with the experimental data.

JAEA Reports

Analyses of the event of control rod withdrawal under low-power operation in "Fugen".

*; *; *

PNC TN9410 88-101, 154 Pages, 1988/07

PNC-TN9410-88-101.pdf:5.33MB

Anticipated reactor transient events under low-power operation have become of interest after the Chernobyl Nuclear Power Plant accident. Therefor, in the present study, the analyses of the event of control rod withdrawal under low-power operation have been carried out for "Fugen", prototype advanced thermal reactor. Analyses have been performed by using EUREKA-2 code, a coupled nuclear thermal hydrodynamic kinetic code. As the result of the analyses, it is confirmed that the effect of coolant void reactivity is almost neglected and that the fuel entarpy is within the limint of established value in the principal guide of reactivity insertion events. In conclusion, there is no matter with control rod withdraw as under low-power operation in design base for "Fugen".

JAEA Reports

Calculation accuracy analysis of core design code for ATR (Advanced thermal reactor) demonstration plant

*; *

PNC TN9410 88-097, 283 Pages, 1988/07

PNC-TN9410-88-097.pdf:20.37MB

The calculation accuracy of the WIMS-ATR/LAIMON-2A codes, or a core design code for the ATR Demonstration Plant was evaluated on the basis of the experimental data from the first to the eighth cycle of the FUGEN. The major results obtained in this study are shown below; (1)The calculation accuracy evaluation based on the data of the type-A fuel cores (from the first to the third cycle). (a)The calculation errors of power distribution are about 3.0% $$sim$$ 3.9% in terms of the RMS (Root Mean Square) of the relative errors for the segment power, and 1.1% $$sim$$ 1.9% in terms of the RMS of the relative errors for the channel power. Therefore, the calculation values are in good agreement with experimental values. (b)The relative calculation errors of the channel power peaking factors are as small as -0.9% $$sim$$ 0.1%, Whereas the relative calculation errors of the hottest segment powers are large as -4.0% $$sim$$ -0.7%. This is caused by the disagreement between measured and calculated axial power distributions. (c)The critieal eigen values of the LAYMON-2A code are 0.982 $$sim$$ 0.992, which are underestimated about 1.8%$$Delta$$k compared with the experimental value Keff=1.0. (d)The relative calculation errors of neutron flux distribution based on the PCM readings are 3.6% $$sim$$ 5.2%. Therefore, the calculation values are relatively in good agreement with experimental values. (e)The calculation accuracy based on WIMS-ATR for power distributions, especially channel power distributions, channel power peaking factors and neutron flux distributions, has been improved in comparison with METHUSELAH-II. However, the calculation errors of the hottest segment powers have been increased. The reason is that the agreement of calculated and measured axial power distributions by WIMS-ATR is deteriorated in comparison with METHUSELAH-II. (2)The calculation accuracy evaluation based on the data of the type-B fuel cores (from the fourth to the eighth cycle). (a)The ...

JAEA Reports

Improvement of calculational accuracy of coolant void reactivity by using WIMS-ATR code

Fukumura, Nobuo*; *; *

PNC TN9410 88-072, 162 Pages, 1988/06

PNC-TN9410-88-072.pdf:11.7MB

Improvement of calculational accuracy of ATR coolant void reactivity has been tried by using the more detailed calculational method as for the derivation of diffusion constant. The former calculational method of deriving the diffusion constant in WIMS-ATR code is the followings: (1)Calculation of collision probability with use of the model dividing the cell into three region (fuel, air gap and moderator). (2)Calculation of diffusion constant according to Bonalmi theory with use of the above calculated collision probability values. The present method is the one which calculates the collision probability more accurately by using multi region model including fuel, coolant, pressure tube and so on. For calculation of collision probability the CLUP code was used. By using the present code with the new calculational model of diffusion constant the coolant void reactivity were calculated. The following results were obtained: (1)The value of the diffusion constant is larger than the former one. This fact is more remarkable in the high voidage core and in the fast neutron energy region. (2)The present calculation accuracies of coolant void reactivity are 0.2%$$Delta$$k for the DCA core and 0.07%$$Delta$$k for the FUGEN core. These values are a half of the former ones. From the above results it is concluded that the present calculation model of the diffusion constant is effective for improvement of calculational accuracy of coolant void reactivity in ATR.

JAEA Reports

Verification of the EUREKA-ATR Code Analysis of the SPERT-III E-core experiment

*; *; *

PNC TN9410 88-057, 167 Pages, 1988/06

PNC-TN9410-88-057.pdf:6.71MB

EUREKA-ATR, a coupled nuclear thermal hydrodynamic multi-dimensional kinetic code, was adapted for the testing of models and methods. Code evaluations were made with the reactivity insertion experiments of the SPERT-III$$cdot$$E-core, a slightly enriched oxide core. The code was tested for non-damaging power excursions including a wide range of initial operating conditions, such as cold-startup, hot-startup and operating-power initial conditions. Comparisons resulted in a good agreement between calculated and experimental power, energy, reactivity and clad surface temperature.

JAEA Reports

Coolant void reactivity estimation of ATR by using gadolinia fuel rod

Fukumura, Nobuo*; *; *

PNC TN9410 88-030, 87 Pages, 1988/02

PNC-TN9410-88-030.pdf:11.3MB

It is a difficult problem to shift coolant void reactivity of ATR to more negative. However it has been clarified by critical experiments with using DCA that use of the fuel pins containing gadolinia is very effective for decreasing coolant void reactivity. Then the analytical study has been done in order to find the solution which has the negative coolant void reactivity of ATR. The lattice is composed of the 36-rod fuel cluster and the lattice pitch is 24.2 cm square. The average enrichment through the MOX fuel cluster is 2.53 or 2.88 w/o Pufis.. This lattice is almost the same one as the demonstration power plant of ATR which is now under designing. The calculation code which is used is WIMS ATR - CITATION code system. The calculational input parameter is as follows; (1)Position of gadolinia poisoned fuel pin. (2)Number of gadolinia poisoned fuel pin. (3)Concentration of gadolinia contained in the fuel pin. (4)Concentration of $$^{10}$$B in the D$$_{2}$$O moderator. (5)0, 20, 40, 60, 80 and 100% coolant void. The burnup calculation is up to 20 GWD/T. This is the core average value. The output of calculation is as follows: (1)Coolant void reactivity. (2)Reactivity coefficient of coolant void. (3)Local power peaking. (4)Decay curve of gadolinia concentration through burn up. According to the present study, the following understanding is clarified: (1)The best solution which has the negative coolant void reactivity value through burn up is the use of the 36-rod MOX fuel cluster which has the gadolinia poisoned fuel pin both in the inner and in the middle array of the cluster. (2)The number of the gadolinia poisoned fuel pin is 3 both in the inner and in the middle array. (3)Concentration of the gadolinia is 5 w/o. It is explained that the above effect is due to decreasigg the thermal neutron shielding of H$$_{2}$$O coolant by the strong neutron absorber gadolinia.

JAEA Reports

MOX fuel utilization in ATR

*; *

PNC TN3410 87-010, 60 Pages, 1987/12

PNC-TN3410-87-010.pdf:1.05MB

ATR, a heavy-water moderated boiling-light-water cooled reactor developed in Japan, is a unique reactor with out-standing flexibility regarding nuclear fuel utilization, because it has superior properties concerning the utilization of plutonium, recovered uranium and depleted uranium. The development of this type of reactor is expected to contribute both to the stable supply of energy and to the establishment of plutonium utilization in Japan. Much effort has been and will be made on the development of Plutonium utilization technology in ATR. which is supported by the experience of MOX fuel utilization in the Fugen and by following R&D. (1)Clarifying characteristics of MOX fueled core on reactor physics and thermohydraulics, especially on reactor physics. (2)Clarifying MOX fuel performance. (3)Development of design, fabrication and inspection of MOX fuel. Fugen, which is the prototype reactor of 165 MWe output using plutonium uranium mixed-oxide(MOX) fuel, has been in commercial operation since March 1979, achieving about 60% of average load factor in more than eight years. The potential of MOX fuel has been demonstrated successfully by the operation of the Fugen. A total of 334 MOX and 342 UO$$_{2}$$ fuel assemblies were loaded into the core during ten refuelling times. The maximum burn-up of fuel has reached 18,500 MWd/t and no fuel has failed.

29 (Records 1-20 displayed on this page)