[en] WIMSTAR (Version 4) is a FORTRAN-IV computer program developed to generate data files for the WIMS lattice code library from the ENDF/B data base. The program must be used in conjunction with the AMPX-II system and has been designed for implementation as a module of that system. This report describes the structure, implementation and use of the AMPX/WIMSTAR system

[en] During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

[en] Group cross sections in the resolved resonance region are commonly computed for each nuclide independently of other resonance nuclides present in the fuel mixture. While this technique is usually entirely adequate for uranium fuel cycles, it is necessary to establish its legitimacy for closed thorium fuel cycles topped with fissile uranium or plutonium by analysis of a number of representative cases. At the same time cross sections originating from WIMS (Winfrith Improved Multigroup Scheme) calculations are compared with values computed in this study. In this context, particular attention is paid to the adequacy of the lower boundary for the WIMS resonance treatment. All calculations are based on heavy nuclide cross sections from the ENDF/B-IV data compilaton (Evaluated Nuclear Data File). Appreciable interaction effects have been determined for all nuclides except for ²³²Th. In most cases, these are due to the strong ²³²Th resonance doublet at 21.8 eV and 23.5 eV but some effects also result from resonances of ²³⁴U (5.19 eV, 48.75 eV), ²³⁶U (5.45 eV), ²⁴²Pu (2.67 eV) and others. The influence of mutual interaction on the infinite lattice multiplicaton factor is very small in comparison to the effects of self-shielding. WIMS cross sections do not always compare well with the values computed in the study, but discrepancies are in most cases related to the different sources of data. Interaction effects are not explicitly taken into account in WIMS. Several nuclides (²³³Pa, ²³³U, ²⁴⁰Pu, ²⁴²Pu) show appreciable self-shielding below the WIMS resonance region and are therefore not treated adequately. The impact of these discrepancies on the multiplication factor is relatively small, however, because of error cancellation. (author)

[en] A simple model was constructed to predict the property transients resulting from the deflagration of a combustible mixture in a sphere or cylinder with venting of the gas mixture to the environment. A computer program VENT, was written to solve the model equation. The model will be particularly useful for studying hydrogen burning effects in loss-of-coolant plus losss of emergency coolant accidents in CANDU reactors

[en] The ORIGEN-S radionuclide generation and depletion code has been used to predict the mass, activity and decay heat of radionuclides produced in Bruce-A UO₂ reactor fuel and the Zircaloy cladding and of radionuclides produced by neutron activation of impurities in both the UO₂ fuel and the Zircaloy cladding. This compilation has been produced to provide initial radionuclide inventory data for use in the safety assessment of the disposal of used CANDU fuel. The methods used to generate the inventories are described. A comparison of the results with those generated using the CANIGEN-II code is also presented

[en] Critical experiments involving a small region of test fuel substituted into a reference lattice have traditionally been analyzed using diffusion codes to extract lattice physics parameters of the test fuel such as the critical buckling and the associated bias in the calculation of k_eff . A method that was first developed in 2006 uses a version of MCNP that was modified to allow the analyst to selectively change fission neutron production in various parts of the model. This paper describes the modification made to MCNP, demonstrates how the substitution experiment analysis is done through several examples using data from the ZED-2 critical facility, and finally, quantifies the expected uncertainties in the method. (author)

[en] As the cost of developing completely new computer codes becomes prohibitive, designers of nuclear facilities are turning to more cost-effective approaches for meeting increasingly strict regulatory requirements applied to safety-related analysis. For designing and licensing the MAPLE family of research reactors, Atomic Energy of Canada Ltd. (AECL) is employing the strategy of adapting major existing codes by linking them together within networks of custom-built interface software. This approach builds on the international investment in developing, maintaining, and verifying existing primary codes and focuses on the less onerous development of interface codes. The resultant code systems are then validated for the new applications of interest

[en] Critical experiments involving a small region of test fuel substituted into a reference lattice have traditionally been analyzed using diffusion codes to extract lattice physics parameters of the test fuel such as the critical buckling and the associated bias in the calculation of k_eff. A method that was first developed in 2006 uses a version of MCNP5 that was modified to allow the analyst to selectively change fission neutron production in various parts of the model. This paper describes the modification made to MCNP5, demonstrates how the substitution experiment analysis is done through several examples using data from the ZED-2 critical facility, and finally, quantifies the expected uncertainties in the method. (author)

[en] Atomic Energy of Canada Limited is building the 10-MW_t MAPLE-X10 reactor facility as a dedicated producer of medical and industrial radioisotopes. Dosimetry aspects of the MAPLE-X10 nuclear design include the calculated thermal and fast neutron flux distributions throughout the reactor assembly and the rate of heat generation in reactor materials and components. Examples of the resolution of design issues are also presented, such as the use of fission counters and ion chambers to provide diverse methods of detecting neutron flux levels and the use of the difference between photon and neutron signals to guard against the effects of downgrading of the heavy-water reflector. Computer codes employed in the calculations include MCNP, ONEDANT, WIMS-AECL, and 3DDT

[en] This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community