No abstract available

[en] The cross sections of many important nuclides are represented in Reich-Moore (RM) formalism in the recent American Evaluated Nuclear Data file, ENDF/B-VI. Processing of cross sections with RM resonance parameters is much more difficult than the other multilevel formalisms such as MLBW and Adler-Adler. In this paper, we derive a rational approximation to the RM collision matrix in the vicinity of a resonance. This simplifies the cross section processing. The energy range of the validity of this approximation in the vicinity of a resonance is also derived. Choosing Ni⁵⁸ as an example, results of our approximation for a non-fissile nuclide are given for two typical s-wave resonances. Our rational approximation method is found to work with good accuracies in the vicinity of resonances

No abstract available

[en] Rational approximation method developed to calculate s-wave collision matrix for non-fissile nuclides in Reich-Moore (RM) resonance formalism is extended to the p-wave part of the RM collision amplitudes

[en] Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed

[en] In order to validate the non-adjusted multigroup cross section set recently prepared from the Japanese Evaluated Nuclear Data Library - Version 2 (JENDL-2), some of the integral parameters of several fast critical benchmark reactor assemblies were analysed by comparing the measured values with those predictions using this cross section set and also by intercomparing the values predicted by various cross section sets. The observations of analyses are presented in this report. The benchmark assemblies have been chosen such that there is a variety in core size, majorfuel component, reflector used etc. One dimensional homogeneous diffusion theory model was used and suitable corrections were added. The assemblies considered were among the ones recommended by the Cross Section Evaluation Working Group (CSEWG), viz. ZPR-3-12, ZPR-3-48, ZPR-3-49, ZPR-3-50, ZPR-3-53, ZPR-3-54, ZPR-3-56B, ZPR-6-7 and ZPR-6-6A. (author). 10 refs., 25 tabs

[en] Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis

[en] The role of the Nuclear Data Section of IAEA has been unique and significant in India's nuclear data related work. Experience at Indira Gandhi Centre for Atomic Research Centre (IGCAR) has shown that the services of NDS-IAEA have been vital in building a capability at our centre, towards preparation of multigroup constants for application in fast reactor physics calculations, starting from the basic evaluated nuclear data libraries. However, bottlenecks exist in realizing completeness in the benefits due to non-availability of certain data or codes to our country. Problems also exist due to lack of manpower and computer resources. In this presentation, our experiences with nuclear data processing, the problems faced etc. are outlined. Some suggestions to NDS-IAEA are also given towards removing some of the problems. (author). 11 refs, 1 fig

[en] A method for generating subgroup or band parameters from resonance tables is described. A computer code SPART was written using this method. This code generates the subgroup parameters for any number of bands within the specified broad groups at different temperatures by reading the required input data from the binary cross section library in the Cadarache format. The results obtained with SPART code for two bands were compared with that obtained from GROUPIE code and a good agreement was obtained. Results of the generation of subgroup parameters in four bands for sample case of ²³⁹Pu from resonance tables of Cadarache Ver.2 library is also presented. 6 refs, 2 tabs

[en] Activation cross sections of Al²⁷, Mn⁵⁵, Fe, Co⁵⁹, Cu, Au¹⁹⁷, Th²³² and U²³⁵ in 100 group structure were calculated from ENDL/84-V for use in estimation of activities of various foils in the proposed PFBR mock up shield experiment at APSARA reactor. Activation cross sections of Zr and its isotopes were also calculated from ENDF/B-VI for activation studies in FBTR. Though cross sections were required for specific reactions for the above purposes, cross sections for all reactions have been calculated and area available with us. 4 refs