ASTM standard recommended guide on application of ENDF/A cross section and uncertainty file: establishment of the file

A new ASTM Standard Recommended Guide on Application of ENDF/A Cross Section and Uncertainty File is in preparation by ASTM Committee E10 on Nuclear Technology and Applications. This ASTM Standard is being prepared in support of the standardization of physics-dosimetry procedures and data needed for Light Water Reactor (LWR) power plant pressure vessel and support structure materials surveillance and test reactor development programs. The main subject of this paper is the estabilishment of the ENDF/A Cross Section and Uncertainty File. The development of evaluated cross section files such as the evaluated nuclear data file, ENDF/B, has occurred mainly to meet the needs of physics calculators. These files are tested by calculations of well-measured benchmark problems such as reactivity or critical mass measurements. Data in the files have then been re-evaluated where disagreements with the benchmark measurements indicate data to be deficient

Bubble Formation: A Bibliography

Bubble phenomena have been given a new meaning with their study in relation to the kinetic behavior of reactors. Prior to their study in relation to physics, the bulk of work on bubble phenomena concerned naval engineering problems of behavior in cavitation and water entry behavior. This bibliography is intended to fill the need of the reactor physicist as well as the naval engineer. An attempt has been made to include all available references on bubble phenomena and associated effects. A subject index has been purposely omitted. It is felt that the breakdown in content headings is sufficient to ascertain areas of interest. There will be overlapping of headings and to find all possible entries, a search through the headings may be desirable. To increase the usefulness of this bibliography the location of an abstract has been cited wherever possible following the reference. Classified reports are included; however, their titles contain no classified information. Sources used in compiling this bibliography are: Chemical Abstracts, Industrial Arts Index, Applied Mechanics Review, Nuclear Science Abstracts, the AEC Abstracts of Classified Literature, the AEC card catalogs available at Atomics International, and the bibliographic services of Armed Services Technical Information agency. (auth

Interlaboratory reaction rate program. 12th progress report, November 1976-October 1979

The Interlaboratory Reaction Rate UILRR) program is establishing the capability to accurately measure neutron-induced reactions and reaction rates for reactor fuels and materials development programs. The goal for the principal fission reactions, /sup 235/U, /sup 238/U and /sup 239/Pu, is an accuracy to within +- 5% at the 95% confidence level. Accurate measurement of other fission and nonfission reactions is also required, but to a lesser accuracy, between +- 5% and 10% at the 95% confidence level. A secondary program objective is improvement in knowledge of the nuclear parameters involved in the standarization of fuels and materials dosimetry measurements of neutron flux, spectra, fluence and burnup

LWR-PV Surveillance Dosimetry Improvement Program review graphics

A primary objective of the multilaboratory program is to prepare an updated and improved set of dosimetry, damage correlation, and the associated reactor analysis ASTM standards for LWR-PV irradiation surveillance programs. Supporting this objective are a series of analytical and experimental validation and calibration studies in Benchmark Neutron Fields, reactor Test Regions, and operating power reactor Surveillance Positions. These studies will establish and certify the precision and accuracy of the measurement and predictive methods which are recommended for use in these standards. Consistent and accurate measurement and data analysis techniques and methods, therefore, will have been developed and validated along with guidelines for required neutron field calculations that are used to (1) correlate changes in material properties with the characteristics of the neutron radiation field and (2) predict pressure vessel steel toughness and embrittlement from power reactor surveillance data

Surveillance dosimetry of operating power plants

The main focus of the research efforts presently underway is the LWR power reactor surveillance program in which metallurgical test specimens of the reactor PV and dosimetry sensors are placed in three or more surveillance capsules at or near the reactor PV inner wall. They are then irradiated in a temperature and neutron flux-spectrum environment as similar as possible to the PV itself for periods of about 1.5 to 15 effective full-power years (EFPY), with removal of the last capsule at a fluence corresponding to the 30- to 40-year plant end-of-life (EOL) fluence. Because the neutron flux level at the surveillance position is greater than at the vessel, the test is accelerated wit respect to the vessel exposure, allowing early assessment of EOL conditions

Spent thermal reactor fuel assembly characterization with solid state tract recorders

A scoping experiment to characterize the neutron field generated from a Light Water Reactor spent fuel assembly has been successfully completed. Solid State Track Recorder (SSTR) neutron dosimeters were exposed at the surface of a spent fuel assembly from a Pressurized Water Reactor. Acceptable track densities were obtained. From these SSTR neutron dosimetry measurements, an absolute neutron flux of about 8 x 10/sup 3/ n/cm/sup 2/ sec was determined at the surface of the spent fuel assembly deduced neutron energy flux, with a mean neutron energy of about 1.3 MeV, is intimately dependent upon the actinide content of the spent fuel

Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

Cross sections required for FMIT dosimetry

The Fusion Materials Irradiation Test (FMIT) facility, currently under construction, is designed to produce a high flux of high energy neutrons for irradiation effects experiments on fusion reactor materials. Characterization of the flux-fluence-spectrum in this rapidly varying neutron field requires adaptation and extension of currently available dosimetry techniques. This characterization will be carried out by a combination of active, passive, and calculational dosimetry. The goal is to provide the experimenter with accurate neutron flux-fluence-spectra at all positions in the test cell. Plans have been completed for a number of experimental dosimetry stations and provision for these facilities has been incorporated into the FMIT design. Overall needs of the FMIT irradiation damage program delineate goal accuracies for dosimetry that, in turn, create new requirements for high energy neutron cross section data. Recommendations based on these needs have been derived for required cross section data and accuracies

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity