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Title: CALCULATION OF TRITIUM RETENTION AND RELEASE FROM COMPONENTS IN GROUT- SEGMENT 6 METALLIC WASTE FROM DEMOLISHED BUILDING 232-F

Abstract

The amount of tritium remaining within and the release rate out of stainless steel process waste from the 232-F Tritium Facility at SRS is calculated as a function of time using the historical exposure of pipe during operation of the facility (1955-1958) and its subsequent deactivation and lay-up. The solution and diffusion of tritium in the wall is the mechanism that governs both the tritium contamination of the pipe during operation and its gradual release after deactivation, including radioactive decay while in the metal. This analysis applies to Segment 6 of the so-called Components in Grout waste form. Results of these calculations will be used in the Groundwater Transport assessment, part of the analysis of the Components in Grout.

Authors:
Publication Date:
Research Org.:
SRS
Sponsoring Org.:
USDOE
OSTI Identifier:
899689
Report Number(s):
WSRC-STI-2007-00051
TRN: US0702103
DOE Contract Number:
DE-AC09-96SR18500
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE WASTES, AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; 36 MATERIALS SCIENCE; CONTAMINATION; DEACTIVATION; DECAY; DIFFUSION; GROUTING; RETENTION; STAINLESS STEELS; TRANSPORT; TRITIUM; WASTE FORMS; WASTES

Citation Formats

Clark, E. CALCULATION OF TRITIUM RETENTION AND RELEASE FROM COMPONENTS IN GROUT- SEGMENT 6 METALLIC WASTE FROM DEMOLISHED BUILDING 232-F. United States: N. p., 2007. Web. doi:10.2172/899689.
Clark, E. CALCULATION OF TRITIUM RETENTION AND RELEASE FROM COMPONENTS IN GROUT- SEGMENT 6 METALLIC WASTE FROM DEMOLISHED BUILDING 232-F. United States. doi:10.2172/899689.
Clark, E. Fri . "CALCULATION OF TRITIUM RETENTION AND RELEASE FROM COMPONENTS IN GROUT- SEGMENT 6 METALLIC WASTE FROM DEMOLISHED BUILDING 232-F". United States. doi:10.2172/899689. https://www.osti.gov/servlets/purl/899689.
@article{osti_899689,
title = {CALCULATION OF TRITIUM RETENTION AND RELEASE FROM COMPONENTS IN GROUT- SEGMENT 6 METALLIC WASTE FROM DEMOLISHED BUILDING 232-F},
author = {Clark, E},
abstractNote = {The amount of tritium remaining within and the release rate out of stainless steel process waste from the 232-F Tritium Facility at SRS is calculated as a function of time using the historical exposure of pipe during operation of the facility (1955-1958) and its subsequent deactivation and lay-up. The solution and diffusion of tritium in the wall is the mechanism that governs both the tritium contamination of the pipe during operation and its gradual release after deactivation, including radioactive decay while in the metal. This analysis applies to Segment 6 of the so-called Components in Grout waste form. Results of these calculations will be used in the Groundwater Transport assessment, part of the analysis of the Components in Grout.},
doi = {10.2172/899689},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Feb 09 00:00:00 EST 2007},
month = {Fri Feb 09 00:00:00 EST 2007}
}

Technical Report:

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  • This report describes an estimate of the release of tritium from contaminated concrete from the demolition of the old 232-F Tritium Facility at the Savannah River Site. The estimate uses data from the scientific literature and information about tritium migration in concrete developed during studies of tritium in concrete at SRS.
  • Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutron flux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered andmore » a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.« less
  • On April 14, 1998, a Pacific Northwest National Laboratory (PNNL) researcher performing work in the Building 324 facility approached facility management and asked if facility management could turn off the tritium sampler in the main exhaust stack. The researcher was demonstrating the feasibility of treating components from dismantled nuclear weapons in a device called a plasma arc furnace and was concerned that the sampler would compromise classified information. B and W Hanford Company (BWHC) operated the facility, and PNNL conducted research as a tenant in the facility. The treatment of 200 components in the furnace would result in the releasemore » of up to about 20 curies of tritium through the facility stack. The exact quantity of tritium was calculated from the manufacturing data for the weapons components and was known to be less than 20 curies. The Notice of Construction (NOC) approved by the Washington State Department of Health (WDOH) had been modified to allow releasing 20 curies of tritium through the stack in support of this research. However, there were irregularities in the way the NOC modification was processed. The researcher was concerned that data performed on the sampler could be used to back-calculate the tritium content of the components, revealing classified information about the design of nuclear weapons. He had discussed this with the PNNZ security organization, and they had told him that data from the sampler would be classified. He was also concerned that if he could not proceed with operation of the plasma arc furnace, the furnace would be damaged. The researcher told BWHC management that the last time the furnace was shut down and restarted it had cost $0.5 million and caused a six month delay in the project`s schedule. He had already begun heating up the furnace before recognizing the security problem and was concerned that stopping the heatup could damage the furnace. The NOC that allowed the research did not have an explicit requirement to operate the sampler during a release. The sampler was installed several years previously for other research. After reviewing the NOC and other safety basis documents, and after consulting environmental compliance specialists, facility management agreed to turn off the sampler.« less
  • Spent targets (Tritium Producing Burnable Absorber Rods, TPBARs) from the Commercial Light Water Reactor-Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site will be sent to waste disposal contained in a so-called ''overpack''. The tritium permeation rate through a welded stainless steel overpack was estimated using a finite difference computer program in a previous report. This report is an evaluation of tritium permeation through three additional overpack designs: (1) a stainless steel overpack sealed using a mechanical closure with a metal gasket, (2) a mild steel overpack with the same mechanical closure, and (3) an aluminum overpack sealed by welding.
  • This report describes an estimate of how much tritium will be held up in those parts of the 232-H process that will remain in the building after deactivation The anticipated state of this tritium is also discussed. This information will be used to assess the radiological status of the deactivated facility.