Criticality problems nuclear data

Calculations of core cell bum-up in the N2 PWR were performed by the spectral code CETERA [16]. The fission product and actinide activities were estimated using the RECOL [17] library data base, which was generated on the bases of the latest versions of the evaluated nuclear data files, ENDF/B-V, with corrections based on the results of critical experiments [18]. The criticality problem was solved for a realistic 3-D geometry model of a TFC by Monte-Carlo with RECOL and checked with MCNP [19] for fresh fuel load. One-group cross-sections were prepared for bum-up calculation of critical loads of both fresh and spent fuel and input to ORIGEN-2 [20] for detailed radionuclide content calculations. 

In the period of 1998-99, two sets of experiments focused on problems of rapid decrease of concentration of boric acid in reactor coolant at nuclear reactor core inlet were performed at the University of Maryland, US, under the auspices of OECD. The situation, when there is an inadvertent supply of boron-deficient water into the reactor vessel, could lead to a rapid (very probably local) increase of reactor core power in reactor, operated at nominal power, or to a start of fission reaction in shut-down reactor (secondary criticality). In the above mentioned experiments the transport of boron-deficient coolant through reactor downcomer and lower plenum was simulated by flow of cold water into a model of reactor vessel. These experiments were selected as the International Standard Problem ISP-43 and organisations, involved in thermal — hydraulic calculations of nuclear reactors, were invited to participate in their computer simulation. Altogether 10 groups took part in this problem with various CFD codes. The participants obtained only data on geometry of the experimental facility, and initial and boundary conditions. 

VII. 14. The SAR should provide sufficient information or references to demonstrate that the computer code, nuclear cross-section data and technique used to complete the criticality safety assessment are adequate. The computer codes used in the safety assessment should be identified and described in the SAR, or adequate references should be included. Verification that the software is performing as expected is important. The SAR should identify or reference all hardware and software (titles, versions, etc.) used in the calculations as weU as pertinent version control information. Correct installation and operation of the computer code and associated data (e.g. cross-sections) should be demonstrated by performing and reporting the results of the sample problems or general validation problems provided with the software package. CapabiUties and limitations of the software that are pertinent to the calculational models should be discussed, with particular attention to discussing limitations that may affect the calculations. 

A systein whose criticality should be diedced is often too complex to End directly applicable experimental measurements. In such a case, computer codes are used to evaluate the criticality safety. The codes rmd data used must be validated by many benchmark calculations. A computer code system named JACS (Ref. 1) has been developed at the Japan Atomic Energy Research Institute (JAE for evaluating nuclear oiticality safety, and more than 700 benclunark calculatiotis, as riiown in Table 1, wm carried out to validate it. Tte computed keff s of the calculations were widely S ead (0.90 to 1.03), and in some cases absolute values of keff woe far from 1.0. Hence, problems are submitted, one on how to improve tim calculation method and another on how to waluate tire criticality safety of a system from the result computed by the cpde. A study has been carried on to answer the latter problem. ... 

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