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DETERMINATION OF THE QUANTITY OF I-135 RELEASED FROM THE AGR-1 TEST FUELS AT THE END OF ATR OPERATING CYCLE 138B

Description: The AGR-1 experiment is a multiple fueled-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and ended with shutdown of the reactor for a brief outage on February 10, 2007 at 0900. The AGR-1 experiment will continue cyclical irradiation for about 2.5 years. In order to allow estimation of the amount of radioiodine released during the first cycle, purge gas flow to all capsules continued for about 4 days after reactor shutdown. The FPMS data acquired during part of that shutdown flow period has been analyzed to elucidate the level of 135I released during the operating cycle.
Date: May 1, 2007
Creator: Hartwell, J. K.; Scates, D. M.; Walter, J. B. & Drigert, M. W.
Partner: UNT Libraries Government Documents Department

Iodine Volatility and pH Control in the AP-600 Reactor

Description: Two design-basis accidents for the AP-600 reactor are formulated and evaluate~ in which significant bypass of the principal pH control system occurs. Some iodine released from the reactor primary system is retained in the Incontainment Refaeling Water Storage Tank (IRWST) water, never entering the containment where trisodium phosphate produces a high pH. Some of this iodine is volatilized and is transported into the reactor containment airspace. in the worst case, a small fraction is released to the environment at design-basis leak rate, yielding a total cumulative iodine release at 30 days of 0.0352 mol (0.023% of core iodine inventory) due to the iodine volatilization bypassing the pH control system. No fission product removal in the containment atmosphere (i.e., natural deposition sprays) is considered.
Date: October 1, 1998
Creator: Beahm, E.C. & Weber, C.F.
Partner: UNT Libraries Government Documents Department

Drying damaged K West fuel elements (Summary of whole element furnace runs 1 through 8)

Description: N Reactor fuel elements stored in the Hanford K Basins were subjected to high temperatures and vacuum conditions to remove water. Results of the first series of whole element furnace tests i.e., Runs 1 through 8 were collected in this summary report. The report focuses on the six tests with breached fuel from the K West Basin which ranged from a simple fracture at the approximate mid-point to severe damage with cladding breaches at the top and bottom ends with axial breaches and fuel loss. Results of the tests are summarized and compared for moisture released during cold vacuum drying, moisture remaining after drying, effects of drying on the fuel element condition, and hydrogen and fission product release.
Date: October 13, 1998
Creator: LAWRENCE, L.A.
Partner: UNT Libraries Government Documents Department

Maximum credible uranium-235 release from 211-H to the ETF

Description: The Effluent Treatment Facility (ETF) criticality review identifies two potential scenarios for criticality: (1) An instantaneous release of a significant quantity of fissile material from a process upset; and (2) A gradual accumulation of fissile material in process vessels. The potential for an instantaneous release from 211-H will be eliminated by the installation of a nuclear safety blank as recommended in the criticality review. With the nuclear safety blank installed, the only mechanism for introducing a gradual accumulation of uranium to the ETF from 211-H will be via Acid Recovery Unit (ARU) and General Purpose (GP) evaporator overheads. This study has determined that the maximum credible annual release of uranium-235 to the ETF from these sources is 106 grams.
Date: September 20, 1988
Creator: Campbell, T. G.
Partner: UNT Libraries Government Documents Department

FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

Description: This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.
Date: December 1, 1997
Creator: Lanning, D.D.; Beyer, C.E. & Painter, C.L.
Partner: UNT Libraries Government Documents Department

Comparison study of AXAIR89Q and AXAIRQ

Description: AXAIR89Q, the primary dose assessment code used at Savannah River Site to predict downwind doses following a short hypothetical atmospheric release, has been improved to incorporate many new features. The new version, AXAIRQ, contains the following improvements: inclusion of dry deposition and the ground shine pathway, 95% dose calculations at user-selected distances, availability of Pasquill-Briggs diffusion coefficients, and user- input mixing height. AXAIRQ can be executed in the same manner as AXAIR89Q by selecting certain inputs. This report shows the differences in committed effective dose equivalents when the new features are invoked for various hypothetical release scenarios.
Date: October 1, 1995
Creator: Simpkins, A.A.
Partner: UNT Libraries Government Documents Department

Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

Description: The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation.
Date: October 21, 1999
Creator: Sindelar, R.L.
Partner: UNT Libraries Government Documents Department

The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

Description: The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.
Date: September 1, 1995
Creator: Myers, B.F.
Partner: UNT Libraries Government Documents Department

Analytical Modeling of Fission Product Releases by Diffusion from Multicoated Fuel Particles

Description: Three levels of fission product diffusional release models are solved exactly. First, the Booth model for a homogeneous uncoated spherical fuel particle is presented and an improved implementation is suggested. Second, the release from a fuel particle with a single barrier layer is derived as a simple alternative to account for a coating layer. Third, the general case of release from a multicoated fuel particle is derived and applied to a TRISO-coated fuel. Previous approaches required approximate numerical solutions for the case of an arbitrary number of coatings with arbitrary diffusivities and arbitrary coating interface conditions.
Date: March 1, 2003
Creator: GELBARD, FRED
Partner: UNT Libraries Government Documents Department

Accident source terms for boiling water reactors with high burnup cores.

Description: The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.
Date: November 1, 2007
Creator: Gauntt, Randall O.; Powers, Dana Auburn & Leonard, Mark Thomas
Partner: UNT Libraries Government Documents Department

Analysis of Radioactive Releases During Proposed Demolition Activities for the 224-U and 224-UA Buildings - Addendum

Description: A post-demolition modeling analysis is conducted that compares during-demolition atmospheric concentration monitoring results with modeling results based on the actual meteorological conditions during the demolition activities. The 224-U and 224-UA Buildings that were located in the U-Plant UO3 complex in the 200 West Area of the Hanford Site were demolished during the summer of 2010. These facilities converted uranyl nitrate hexahydrate (UNH), a product of Hanford’s Plutonium-Uranium Extraction (PUREX) Plant, into uranium trioxide (UO3). This report is an addendum to a pre-demolition emission analysis and air dispersion modeling effort that was conducted for proposed demolition activities for these structures.
Date: December 21, 2010
Creator: Napier, Bruce A.; Rishel, Jeremy P.; Droppo, James G.; Joyce, Kevin E. & Strom, Daniel J.
Partner: UNT Libraries Government Documents Department

Synthesis of VERCORS and Phebus data in severe accident codes and applications.

Description: The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, testing and incorporation of improved radionuclide release models into the MELCOR severe accident code.
Date: April 1, 2010
Creator: Gauntt, Randall O.
Partner: UNT Libraries Government Documents Department

Severe accident progression perspectives based on IPE results

Description: Accident progression perspectives were gathered from the level 2 PRA analyses (the analysis of the accident after core damage has occurred involving the containment performance and the radionuclide release from the containment) described in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were obtained. Complete results are discussed in NUREG-1560 and summarized here.
Date: August 1, 1996
Creator: Lehner, J.R.; Lin, C.C.; Pratt, W.T. & Drouin, M.
Partner: UNT Libraries Government Documents Department

Atmospheric relative concentrations in building wakes

Description: This report documents the ARCON96 computer code developed for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation for potential use in control room habitability assessments. It includes a user`s guide to the code, a description of the technical basis for the code, and a programmer`s guide to the code. The ARCON96 code uses hourly meteorological data and recently developed methods for estimating dispersion in the vicinity of buildings to calculate relative concentrations at control room air intakes that would be exceeded no more than five percent of the time. The concentrations are calculated for averaging periods ranging from one hour to 30 days in duration. ARCON96 is a revised version of ARCON95, which was developed for the NRC Office of Nuclear Regulatory Research. Changes in the code permit users to simulate releases from area sources as well as point sources. The method of averaging concentrations for periods longer than 2 hours has also been changed. The change in averaging procedures increases relative concentrations for these averaging periods. In general, the increase in concentrations is less than a factor of two. The increase is greatest for relatively short averaging periods, for example 0 to 8 hours and diminishes as the duration of the averaging period increases.
Date: May 1997
Creator: Ramsdell, J. V., Jr. & Simonen, C. A.
Partner: UNT Libraries Government Documents Department

An evaluation of the suitability of laser-induced fluorescence for measurements of fission-product iodine sorptivity in the MHTGR [modular high-temperature gas-cooled reactor]

Description: Experiments and calculations indicate that laser-induced fluorescence (LIF) lacks the sensitivity needed for sorptivity measurements of I{sub 2} or other molecular species at partial pressures below 10{sup {minus}11} atm. Although the technique may have sufficient sensitivity for measurements of atomic species, the species of interest are, in all likelihood, not atomic. Methods of measurement which would allow the determination of species are proposed. 9 refs., 6 figs.
Date: July 1989
Creator: Sherrow, S. A.
Partner: UNT Libraries Government Documents Department

Containment performance perspectives based on IPE results

Description: Perspectives on Containment Performance were obtained from the accident progression analyses, i.e. level 2 PRA analyses, found in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were gathered. The results summarized here are discussed in detail in volumes 1 and 2 of NUREG 1560. 3 refs., 4 figs.
Date: 1997
Creator: Lehner, J. R.; Lin, C. C. & Pratt, W. T.
Partner: UNT Libraries Government Documents Department

Effect of recrystallization in high-burnup UO{sub 2} on gas release during RIA-type transients

Description: The authors recently proposed a model for irradiation-induced recrystallization (grain subdivision) and swelling in UO{sub 2} fuels wherein the stored energy in the material is concentrated in a network of sink-like nuclei that diminish with dose due to interaction with radiation-produced defects. It is of considerable interest to explore the consequences of recrystallization on gas release during a reactivity initiated accident (RIA). In the absence of recrystallization, gas release during RIA-type transients is generally limited to gas available on grain boundaries and edges due to the very short heatup times (milliseconds), short cooldown periods (seconds), and relatively long intragranular diffusion distances (on the order of micrometers). However, recrystallization provides grain-boundary surfaces that are substantially closer to the gas retained in the bulk material, and thus the potential for much higher gas release. The authors show the calculated burnup at which grain subdivision will occur as a function of fractional radius and fuel temperature for a generic pressurized water reactor irradiation. The FASTGRASS code was used to calculate fission gas behavior during in-reactor irradiation and during the RIA-type transient. Results are given. It is clear from these results that recrystallization of high-burnup UO{sub 2} has implications for the potential consequences of severe accident scenarios such as the RIA type.
Date: October 1, 1994
Creator: Rest, J. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department