[en] Collision probability methods using rational approximations for space shielding of cross sections in nuclear reactor codes allowing for one level of heterogeneity are first reviewed. A Dancoff factor for a cylindrical HTGR fuel pin with grain structure is then derived. A new simple method for cross section space shielding in a doubly heterogeneous HTGR system using collision probabilities and rational approximations is subsequently presented and calculations based on it are compared with calculations based on more elaborate methods. 39 references

No abstract available

[en] A new method for space shielding cross sections in doubly heterogeneous reactor cores is presented. It operates on detailed pointwise or ultrafine group cross sections and can be implemented only in computer programs which generate cross sections at such a fine energy level. This method is accurate and can serve as reference for testing approximate methods of cross-section space shielding

[en] Delayed beta and gamma group-energy spectra were calculated for 180 individual fission-product nuclides having spectral data in ENDF/B-IV fission-product files. The beta spectra, in uniform-grid 75-group structure, and the gamma spectra, in uniform-grid 150-group structure between 0 and 7.5 MeV, are sufficiently detailed for collapsing to any desired practical group structure. These spectra have been used in decay-heat summation calculations for comparisons with corresponding experimental beta and gamma spectra

[en] A simple yet highly accurate method for approximately calculating spectrum-averaged beta energies and beta spectra for radioactive nuclei is presented. This method should prove useful for users who wish to obtain accurate answers without complicated calculations of Fermi functions, complex gamma functions, and time-consuming numerical integrations as required by the more exact theoretical expressions. Therefore, this method should be a good time-saving alternative for investigators who need to make calculations involving large numbers of nuclei (e.g., fission products) as well as for occasional users interested in restricted number of nuclides. The average beta-energy values calculated by this method differ from those calculated by ''exact'' methods by no more than 1 percent for nuclides with atomic numbers in the 20 to 100 range and which emit betas of energies up to approximately 8 MeV. These include all fission products and the actinides. The beta-energy spectra calculated by the present method are also of the same quality

[en] An automated interface between EPRI-CELL, a reactor cell analysis code, and EPRI-CINDER, a point depletion and fission-product code, has been generated for use in the Advanced Recycle Methodology Program (ARMP) system. This enables an efficient sequential operation of the two codes without direct handling of intermediate CELL output that is used to modify the CINDER input at each depletion time step. This alternative to the existing CELL internal depletion capability provides means for new and better application of these codes to wider areas of nuclear problems. The stand-alone EPRI-CINDER code has also been provided with new capabilities and has been modified to accept a simpler optional input structure

[en] The objective of this project has been the production of a data library suitable for calculating the buildup of fission product nuclides during the operation of a thermal power reactor. This has been accomplished by reducing the fission product data from the fourth version of the national reference nuclear data base--ENDF/B into a series of linearized decay chains and calculating the effective yields and cross sections of the relevant nuclides. Two versions of the fission product library have been prepared: an 84 chain master library and a reduced 12 chain library, both of which can be used as input for the computer program CINDER. A users manual for an upgraded version of the burnup program CINDER (renamed EPRI-CINDER) is presented

No abstract available

[en] The report discusses methods used at LASL for calculating group cross sections for doubly heterogeneous HTGR systems of the General Atomic design. These cross sections have been used for the neutronic safety analysis calculations of such HTGR systems at various points in reactor lifetime (e.g., beginning-of-life, end-of-equilibrium cycle). They were also compared with supplied General Atomic cross sections generated with General Atomic codes. The overall agreement between the LASL and the GA cross sections has been satisfactory

[en] FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures