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A compendium of dosimetry cross sections is presented for use in the characterization of fission reactor spectrum and fluence. The contents of this cross section library are based upon the ENDF/B-VI and IRDF-90 cross section libraries and are recommended as a replacement for the DOSCROS84 multigroup library that is widely used by the dosimetry community. Documentation is provided on the rationale for the choice of the cross sections selected for inclusion in this library and on the uncertainty and variation in cross sections presented by state-of-the-art evaluations. MASTEB Acknowledgment The authorsthankDavid W. Vehar,supervisorof the Radiation Metrology La0oratory,for consultation and advice in makingthe formatandcontentof this cross section compendium supportboth the dosimetrist and the reactor experimenter. We also thank CharlesV. Holm for providingthe highest qualityfoil activities and experimental supportfor fielding dosimetry sensors. High qualitydosimetry sensor readingswere critical in allowing the authorsto judge the consistency and accuracyof the cross sections provided in this compendium.

The neutron spectrum characteristics of the primary reactor environments are defined for use by facility customers and to provide an audit trail in support of current quality assurance initiatives. The neutron and gamma environments in the four primary customer environments at SPR-III and ACRR facilities are characterized in detail. Enough detail is provided on other frequently-used environments to support the definition of the 3-MeV and 1-MeV(Si) fluence provided on the Radiation Metrology Laboratory dosimetry reports. Details are provided to enable customers to compare new dosimetry results in the four primary environments with dosimetry reports from previous irradiations. SAND-II and MANIPULATE output listings used to derive the new spectrum-averaged integral parameters are provided in appendices that appear in volume 2 of this document. Intentionally Left Blank _ NATIONAL IJLBOR_TORIES Radiation _trology Laboratory Nuclear Facilitie| and Diagnostic| Department Albuquerque, NN 87185-1142 (505) 845-3250 Sulfur Dosimetry I-JAN-94 llslTt25 Ex_orlmonters J. Doe Organizationz AAA FacilitFt Uncharecterlzed Shot Number: 9181 8hot Datez 01-JAN-1994 16s00 File s UD: [TEST. DATA] SOL_datb_03622. dat LaSt odltod: I6-FEB-94 12s56s25 Counter 1 Background 1.50 +/-0.22 cpm >3 Mov/fac_tots 0.0000 +/-0.00% l-MeV(Sl)/f&c_totz 0.0000 +/-0.00% C£-252/fac_tots 1.0000 +/-0.00%

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Sandia National Laboratories, in the process of characterizing the neutron environments at its reactor facilities, has developed an enhanced version of W. McElroy's original SAND-II code. The enhanced input, output, and plotting interfaces make the code much easier to use. The basic physics and operation of the code remain unchanged. Important code enhancements include the interfaces to the latest ENDF/B-VI and IRDF-90 dosimetry-quality cross sections and the ability to use silicon displacement-sensitive devices as dosimetry sensors.