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Tamaki, Hitoshi; Kimoto, Tatsuya*; Hamaguchi, Yoshikane*; Yoshida, Kazuo
Nihon Genshiryoku Gakkai Wabun Rombunshi, 9(1), p.40 - 51, 2010/03
A criticality accident in MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory, plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accident. Control failures of at least two parameters are needed for occurrence of criticality accident. To evaluate probability of such control failure, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planed to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue to evaluate likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with taking these issues into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method were also shown through a trial application to a model MOX fabrication facility.
Takeda, Toshikazu*; *; Kitada, Takanori*; *
PNC TJ9605 97-001, 100 Pages, 1997/03
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,
Takeda, Toshikazu*; Kitada, Takanori*; *; *
PNC TJ9605 96-001, 120 Pages, 1996/03
This report is composed of the following three parts. (I)Improvement of Calculational Scheme of Reaction Rate Distribution by Monte Carlo Method. (II)Calculational Method for Doppler Reactivity Worth by Using Ultra-Fine Energy Group. (III)Improvement of Calculational Method of Transverse Leakage on Three Dimensional Nodal Transport Code for Hexagonal-Z Geometry.
