KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with fissionable material. In general, continuous-energy Monte Carlo methods are preferred because such an approach avoids many of the assumptions inherent in the multigroup treatment. On the other hand, continuous-energy treatment is much more demanding in terms of computer storage space for data, memory requirements, and calculation speed. Continuous-energy versions of KENO V.a and KENO-VI have been created and are being extensively tested. Generation of ENDF/B-VI continuous-energy cross sections is explained, and the results of the validation and verification of the codes and the data are presented.

KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with fissionable material. In general, continuous-energy Monte Carlo methods are preferred because such an approach avoids many of the assumptions inherent in the multigroup treatment. On the other hand, continuous-energy treatment is much more demanding in terms of computer storage space for data, memory requirements, and calculation speed. Continuous-energy versions of KENO V.a and KENO-VI have been created and are being extensively tested. Generation of ENDF/B-VI continuous-energy cross sections is explained, and the results of the validation and verification of the codes and the data are presented.

The AMPX [1] code system is a modular system of FORTRAN computer programs that relate to nuclear analysis with a primary emphasis on tasks associated with the production and use of multi group and continuous energy cross sections. The module PUFF-III within this code system handles the creation of multi group covariance data from ENDF information. The resulting covariances are saved in COVERX format [2]. We recently expanded the capabilities of PUFF-III to include full handling of covariance data in the resonance region (resolved as well as unresolved). The new program handles all resonance covariance formats in File 32 except for the long-range covariance sub sections. The new program has been named PUFF-IV. To our knowledge, PUFF-IV is the first processing code that can address both the new ENDF format for resolved resonance parameters and the new ENDF 'compact' covariance format. The existing code base was rewritten in Fortran 90 to allow for a more modular design. Results are identical between the new and old versions within rounding errors, where applicable. Automatic test cases have been added to ensure that consistent results are generated across computer systems. (authors)

We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VIL0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes, based on experimental data and theory predictions.The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th; Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, Li-6, B-10, An and for U-235,U-238 fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC): (3) Improved thermal neutron scattering:, (4) An extensive set of neutron cross sections on fission products developed through a WPEG collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron-and proton-induced evaluations up to 150 MeV: (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra:, (9) New radioactive decay data:, (10) New methods for uncertainties and covariances, together with covariance evaluations for some sample cases; and (11) New actinide fission energy deposition.The paper provides an overview of this library; consisting of 14 sublibraries in the same ENDF-6 format as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The Ion-standing underprediction of low enriched uranium thermal assemblies is removed; (b) The U-238 and Pb-208 0 U and Ph reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good agreement for simulations of thermal high-enriched uranium assemblies is preserved; (d) The underprediction of fast criticality of U-233,U-235 and Pu-239 assemblies is removed; and (e) The intermediate spectrum critical assemblies are predicted more accurately.We anticipate that the new library will play an important role in nuclear technology applications, including transport simulations supporting national security, nonproliferation, advanced reactor and fuel cycle concepts, criticality safety, fusion, medicine, space applications, nuclear astrophysics, and nuclear physics facility design. The ENDF/B-VII.0 library is archived at the National Nuclear Data Center, BNL, and can be retrieved from www.nndc.bnl.gov.

The AMPX code system has been limited to processing Version 5 formats of the Evaluated Nuclear Data File (ENDF). Work is in progress to upgrade the AMPX code system to process ENDF/B-VI data. A substantial amount of code development is complete, and the latest version of AMPX has the capability to (a) process cross sections in the resolved resonance region (RRR) using the multipole formalism and (b) produce continuous-energy cross sections from ENDF/B-VI data. Comprehensive testing of the new modules is underway, and the purpose of this work is to establish the capabilities of the new AMPX code system for criticality safety applications.

The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCAMPI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library is designated BUGLE-93. The processing methodology is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, “pseudo problem-independent” format and then collapsed into the final broad-group format. The fine-group library is designated VITAMIN-B6. An extensive integral data testing effort was also performed. In general, results using the new data show significant improvements relative to earlier ENDF data.

Accepted for publication July 31, 1959. The authors are in the Section of Anesthesiology, Yale University School of Medicine, and the Department of Anesthesia, Grace-New Haven Community Hospital, New Haven, Connecticut.

Epinephrine and norepinephrine are biologically active substances which are normally reflexly released in response to a number of stressful situations including hypoxia,7'' " hypercarbia,'7"' hemorrhage, ",7" and general anesthesia.'""" In the process of evaluating the effects of such stressful situations on the amount of oxygen in tissue and available for normal cellular respiration and metabolism, it became necessary to differentiate between those changes in tissue oxygenation which were the result of the stressful situations per se, and those changes in tissue oxygenation which were the result of the reflex release of epinephrine and norepinephrine occasioned by the stressful situations. The present study was instituted to clarify this problem by measuring tissue oxygen concentration (in terms of tissue oxygen tension) in normal adults given infusions of epinephrine and norepinephrine at a time when they were breathing room air at rest. Previous reports on this subject have dealt only with dogs and have been preliminary in nature.9 Polarographic techniques were used. Although they have certain definite and inherent limitations,"' when these limitations are respected, such techniques provide valuable information presently unobtainable in any other manner. The present data relate only to changes in skin oxygen tension.

MACKAY, FRANCES J. M.D.; HEHRE, FREDERICK W. M.D.; GREENE, NICHOLAS M. M.D. Author Information

KENO V.a and KENO VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation and are used for performing criticality calculations of systems with fissionable material. As part of current research at Oak Ridge National Laboratory, continuous- energy or pointwise versions of the KENO codes are being developed for implementation in the SCALE system. Moreover, continuous-energy cross-section processing and transport procedures have been developed to support the Point KENO development. As part of the continuous-energy cross-section development, new procedures based on pointwise collision kinematics have been developed and implemented in the KENO codes. In addition, AMPX cross-section processing modules have been developed for processing ENDF/B evaluations and generating continuous- energy cross-section data for the pointwise versions of KENO. For this work, 50 ENDF/B-VI Release 7 evaluations have been processed to generate a continuous-energy KENO test library. Using the test cross-section library, the continuous-energy transport procedures have been tested by calculating various test problems and comparing with MCNP4C. The details of the continuous-energy transport development and comparison calculations with MCNP are presented in this paper.