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[Dispersion of fission products]

Description: This historic report is on letter which details the research into atmospheric dispersion of fissioviable materials in the event of a reactor accident at Savannah River Plant. Graphs are included.
Date: October 29, 1956
Creator: Menegus, R. L.
Partner: UNT Libraries Government Documents Department

Monthly report of the Design Analysis Group for September 1954

Description: The following topics were discussed in this report: pressurization of the rear face for existing reactors; loss of steam in existing reactors; in-pile boiling in the 105-KER recirculation facilities; feasibility report for special study reactor plant; shielding requirements of the special study reactor plant; sulfuric acid addition to 100-K; radiation from activated iron in a recirculating system; ion exchanger activity from Fe corrosion; thermal shock-KER loop; shielding of tube bundles; adjustment of raw water pH; and Zircaloy fuel element jackets.
Date: October 26, 1954
Creator: Andersen, R. K.
Partner: UNT Libraries Government Documents Department

Estimating pressurized water reactor decommissioning costs: A user`s manual for the PWR Cost Estimating Computer Program (CECP) software. Draft report for comment

Description: With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.
Date: October 1, 1993
Creator: Bierschbach, M. C. & Mencinsky, G. J.
Partner: UNT Libraries Government Documents Department

Safety questions relevant to nuclear thermal propulsion

Description: Nuclear propulsion is necessary for successful Mars exploration to enhance crew safety and reduce mission costs. Safety concerns are considered by some to be an implements to the use of nuclear thermal rockets for these missions. Therefore, an assessment was made of the various types of possible accident conditions that might occur and whether design or operational solutions exist. With the previous work on the NERVA nuclear rocket, most of the issues have been addressed in some detail. Thus, a large data base exist to use in an agreement. The assessment includes evaluating both ground, launch, space operations and disposal conditions. The conclusion is that design and operational solutions do exist for the safe use of nuclear thermal rockets and that both the environment and crews be protected against harmful radiation. Further, it is concluded that the use of nuclear thermal propulsion will reduce the radiation and mission risks to the Mars crews.
Date: October 15, 1991
Creator: Buden, D.
Partner: UNT Libraries Government Documents Department

Thermal analysis of the FSP-1RR irradiation test

Description: The thermal analysis of four unirradiated fuel pins to be tested in the FSP-1RR fuels irradiation experiment was completed. This test is a follow-on experiment in the series of fuel pin irradiation tests conducted by the SP-100 Program in the Fast Flux Test Facility. One of the pins contains several meltwire temperature monitors within the fuel and the Li annulus. A post-irradiation examination will verify the accuracy of the pre-irradiation thermal analysis. The purpose of the pre-irradiation analysis was to determine the appropriate insulating gap gas compositions required to provide the design goal cladding operating temperatures and to ensure that the meltwire temperature ranges in the temperature monitored pin bracket peak irradiation temperatures. This paper discusses the methodology and summarizes the results of the analysis.
Date: October 14, 1992
Creator: Webb, R. H. & Lyon, W. F. III
Partner: UNT Libraries Government Documents Department

Small low mass advanced PBR`s for propulsion

Description: The advanced Particle Bed Reactor (PBR) to be described in this paper is characterized by relatively low power, and low cost, while still maintaining competition values for thrust/weight, specific impulse and operating times. In order to retain competitive values for the thrust/weight ratio while reducing the reactor size, it is necessary to change the basic reactor layout, by incorporating new concepts. The new reactor design concept is termed SIRIUS (Small Lightweight Reactor Integral Propulsion System). The following modifications are proposed for the reactor design to be discussed in this paper: Pre-heater (U-235 included in Moderator); Hy-C (Hydride/De-hydride for Reactor Control); Afterburner (U-235 impregnated into Hot Frit); and Hy-S (Hydride Spike Inside Hot Frit). Each of the modifications will be briefly discussed below, with benefits, technical issues, design approach, and risk levels addressed. The paper discusses conceptual assumptions, feasibility analysis, mass estimates, and information needs.
Date: October 1, 1993
Creator: Powell, J. R.; Todosow, M. & Ludewig, H.
Partner: UNT Libraries Government Documents Department

MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

Description: MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data.
Date: October 1, 1993
Creator: Tautges, T. J.
Partner: UNT Libraries Government Documents Department

Design report: SCDAP/RELAP5 reflood oxidation model

Description: Current SCDAP/RELAP5 oxidation models have proven to under-predict oxidation, and therefore hydrogen production, when modeling reflood during in-pile tests. As an example, while OECD LOFT Experiment LP-FP-2 shows significant increases in temperature and pressure during reflood due to increased oxidation, only minimal additional oxidation is currently predicted with SCDAP/RELAP5. Since SCDAP/RELAP5 predicts a steam rich environment during reflood, the parameter limiting oxidation must be the availability of zircaloy. Two phenomena, not currently modeled, may provide the necessary unoxidized zircaloy during reflood: (1) localized steam starvation prior to reflood, caused by debris blockage or hydrogen generation, or (2) shattering of oxidized cladding during reflood. The objective of this design report is to develop new models to accurately predict zircaloy cladding oxidation during the temperature transients prior to and during reflood. Evidence compiled from postirradiation examination (PIE) of fuel bundles subjected to severe accident conditions from several in-pile tests is used to identify mechanisms for additional cladding oxidation during reflood and to develop specific criteria to determine when these mechanisms are applicable.
Date: October 1, 1992
Creator: Coryell, E. W.; Chavez, S. A.; Davis, K. L. & Mortensen, M. H.
Partner: UNT Libraries Government Documents Department

Nuclear Regulatory Commission Issuances

Description: This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors Decisions (DD), and the Denials of Petitions for Rulemaking (DPRM). The summaries and headnotes preceding the opinions reported herein are not to be deemed a part of those opinions or have any independent legal significance.
Date: October 1, 1994
Partner: UNT Libraries Government Documents Department

Uses of zero-one sampling in probabilistic risk assessment

Description: The recent NUREG-1150 studies and the LaSalle Probabilistic Risk Assessment (PRA) include the most in-depth uncertainty analyses ever performed for commercial nuclear reactors. As a result, the methods used in these studies are often emulated and referenced as being the definitive approach for performing such uncertainty analyses. While the methods are believed to be robust and to reasonably reflect the magnitude of the uncertainties, it is important for future users of these methods to understand some of the subtle points of the analysis. In particular, zero-one sampling (ZOS) was a technique employed extensively in these studies. The purpose of this paper is to clarify the actual use of zero-one sampling in NUREG-1150 and discuss more precisely those applications for which zero-one sampling is appropriate.
Date: October 1, 1993
Creator: Camp, A. L.
Partner: UNT Libraries Government Documents Department

Aging assessment of the boiling-water reactor (BWR) standby liquid control system. Phase 1

Description: Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC system were reviewed to characterize chemical properties and corrosion characteristics of borated solutions. The leading aging degradation concern to date appears to be setpoint drift in relief valves, which has been discovered during routine surveillance and is thought to be caused by mechanical wear. Degradation was also observed in pump seals and internal valves. In general, however, the results of the Phase I study suggest that age-related degradation of SLC systems has not been serious.
Date: October 1, 1992
Creator: Orton, R. D.; Johnson, A. B.; Buckley, G. D. & Larson, L. L.
Partner: UNT Libraries Government Documents Department

Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 3, Development documentation

Description: The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause events at nuclear power plants. This document, Volume III, is a detailed documentation of the development effort and the pilot training program.
Date: October 1, 1993
Creator: Paradies, M.; Unger, L.; Haas, P. & Terranova, M.
Partner: UNT Libraries Government Documents Department

Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 1, Summary

Description: The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause events at nuclear power plants. This document, Volume I is a concise description of the need for the human performance investigation process, the process` components, the methods used to develop the process, the methods proposed to test the process, and conclusions on the process` usefulness.
Date: October 1, 1993
Creator: Paradies, M.; Unger, L.; Haas, P. & Terranova, M.
Partner: UNT Libraries Government Documents Department

Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 2, Investigators`s Manual

Description: The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause event at nuclear power plants. This document, Volume II, is a field manual for use by investigators when performing event investigations. Volume II includes the HPIP Procedure, the HPIP Modules, and Appendices that provide extensive documentation of each investigation technique.
Date: October 1, 1993
Creator: Paradies, M.; Unger, L.; Haas, P. & Terranova, M.
Partner: UNT Libraries Government Documents Department

Fatigue of carbon and low-alloy steels in LWR environments

Description: Fatigue tests have been conducted on A106-Gr B carbon steel and A533-Gr B low-alloy steel to evaluate the effects of an oxygenated-water environment on the fatigue life of these steels. For both steels, environmental effects are modest in PWR water at all strain rates. Fatigue data in oxygenated water confirm the strong dependence of fatigue life on dissolved oxygen (DO) and strain rate. The effect of strain rate on fatigue life saturates at some low value, e.g., between 0.0004 and 0.001%/s in oxygenated water with {approximately}0.8 ppm DO. The data suggest that the saturation value of strain rate may vary with DO and sulfur content of the steel. Although the cyclic stress-strain and cyclic-hardening behavior of carbon and low-alloy steels is distinctly different, the degradation of fatigue life of these two steels with comparable sulfur levels is similar. The carbon steel exhibits pronounced dynamic strain aging, whereas strain-aging effects are modest in the low-alloy steel. Environmental effects on nucleation of fatigue crack have also been investigated. The results suggest that the high-temperature oxygenated water has little or no effect on crack nucleation.
Date: October 1, 1993
Creator: Chopra, O. K.; Michaud, W. F. & Shack, W. J.
Partner: UNT Libraries Government Documents Department

MELCOR 1.8.2 Assessment: IET direct containment heating tests

Description: MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze several of the IET direct containment heating experiments done at 1:10 linear scale in the Surtsey test facility at Sandia and at 1:40 linear scale in the corium-water thermal interactions (CWTI) COREXIT test facility at Argonne National Laboratory. These MELCOR calculations were done as an open post-test study, with both the experimental data and CONTAIN results available to guide the selection of code input. Basecase MELCOR results are compared to test data in order to evaluate the new HPME DCH model recently added in MELCOR version 1.8.2. The effect of various user-input parameters in the HPME model, which define both the initial debris source and the subsequent debris interaction, were investigated in sensitivity studies. In addition, several other non-default input modelling changes involving other MELCOR code packages were required in our IET assessment analyses in order to reproduce the observed experiment behavior. Several calculations were done to identify whether any numeric effects exist in our DCH IET assessment analyses.
Date: October 1, 1993
Creator: Kmetyk, L. N.
Partner: UNT Libraries Government Documents Department

The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

Description: The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.
Date: October 26, 1992
Creator: Lancaster, D.
Partner: UNT Libraries Government Documents Department

SPACE-R thermionic space nuclear power system: Design and technology demonstration. Task 1.5.6, Moderator containment laboratory experiment test plant (CDRL No. 5)

Description: The preferred moderator being considered for SPACE-R is yttrium hydride encased in beryllium tubes. The baseline beryllium performs a dual function as it acts as a moderator and provides containment for hydrogen. The permeation rate of hydrogen from the hydride through the beryllium shell at the operating temperature is an important factor for the functionality and reliability of the Be-YHx moderator. Hydrogen containment capability of beryllium is comparable to enamel which was used in SNAP and Topaz II reactors. However, limited experimental data base exists for the hydrogen permeation through fabricated beryllium enclosures at high temperature. Permeation of hydrogen in beryllium is strongly affected by surface conditions, thickness of surface oxide, surface and bulk traps, impurity content and microstructure. The objective of this experiment is to determine the permeation rate of hydrogen from yttrium hydride and zirconium hydride through beryllium in the temperature range of 773 K--973 K. In addition, Topaz II type zirconium hydride specimens with and without the proprietary oxide coating canned in stainless steel will be tested to measure the hydrogen permeation rate. The TSET SS-canned ZrHx samples currently at Phillips Laboratory will be used for the latter test with Phillips Laboratory participation at the SPI hydrogen leak test stand. A key technology demonstration of the effectiveness of transferred arc plasma spraying of a 1 mil Molybdenum coating on the Be cladding will be performed. The effectiveness of the Molybdenum coating in preventing any interaction of Be with Stainless Steel in NaK will be assessed and demonstrated.
Date: October 1, 1993
Partner: UNT Libraries Government Documents Department

Mechanical-property degradation of cast stainless steel components from the Shippingport reactor

Description: The mechanical properties of cast stainless steels from the Shippingport reactor have been characterized. Baseline properties for unaged materials were obtained from tests on either recovery-annealed material or material from a cooler region of the component. The materials exhibited modest decrease in impact energy and fracture toughness and a small increase in tensile strength. The fracture toughness J-R curve, J{sub IC} value, tensile flow stress, and Charpy-impact energy of the materials showed very good agreement with estimations based on accelerated laboratory aging studies. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy that would be achieved after long-term aging, were established from materials that were aged further in the laboratory at temperatures between 320 and 400{degrees}C. The results showed very good agreement with estimates; the activation energies ranged from 125 to 250 kJ/mole and the minimum room temperature impact energy was <75 J/cm{sup 2}. The estimated impact energy and fracture toughness J-R curve for materials from the Ringhals reactor hot and crossover-leg elbows are also presented.
Date: October 1, 1991
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Integral Fast Reactor Program. Annual progress report, FY 1993

Description: This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.
Date: October 1, 1994
Creator: Chang, Y. I.; Walters, L. C.; Laidler, J. J.; Pedersen, D. R.; Wade, D. C. & Lineberry, M. J.
Partner: UNT Libraries Government Documents Department

Estimation of mechanical properties of cast stainless steels during thermal aging in LWR systems

Description: A procedure and correlations are presented for predicting Charpy- impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of aged cast stainless steels from known material information. The ``saturation`` impact strength and fracture toughness of a specific cast stainless steel, i.e., the minimum value that would be achieved for the material after long-term service, is estimated from the chemical composition of the steel. Mechanical properties as a function of time and temperature of reactor service are estimated from impact energy and flow stress of the unaged material and the kinetics of embrittlement, which are also determined from chemical composition. The J{sub IC} values are determined from the estimated J-R curve and flow stress. Examples of estimating mechanical properties of cast stainless steel components during reactor service are presented. A common ``lower-bound`` J-R curve for cast stainless steels of unknown chemical composition is also defined for a given grade of steel, ferrite content, and temperature.
Date: October 1, 1991
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Applicability of trends in nuclear safety analysis to space nuclear power systems

Description: A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication.
Date: October 1, 1992
Creator: Bari, R. A.
Partner: UNT Libraries Government Documents Department

Assessment of the use of H{sub 2}, CH{sub 4}, NH{sub 3} and CO{sub 2} as NTP propellants. Revision

Description: In this paper the effect of changing from the traditional NTP coolant, hydrogen, to several alternative coolant is studied. Hydrogen is generally chosen as an NTP coolant, since its use maximizes the specific impulse for a given operating temperature. However, there are situations in which it may not be available as optional. The alternative coolant which were considered are ammonia, urethane, carbon dioxide and carbon monoxide. A particle bed reactor (PBR) generating 200 MW and coolant by hydrogen was used as the baseline against which all the comparisons were made. Both 19 and 37 element cases were considered and the large number of elements was found to be necessary in the case of the carbon monoxide. The coolant reactivity worth was found to be directly proportional to the hydrogen coolant content. It was found that due to differences in the thermophysical proportions of the coolant that it would not be possible to use one reactor for all the coolants. The reactor would have to constructed specifically for a coolant type.
Date: October 1, 1991
Creator: Selcow, E. C.; Davis, R. E.; Perkins, K. R.; Ludewig, H. & Cerbone, R. J.
Partner: UNT Libraries Government Documents Department