ML051040346
| ML051040346 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 05/06/2005 |
| From: | Gillen D NRC/NMSS/DWMEP/DD |
| To: | Jesse Rollins Yankee Atomic Electric Co |
| Hickman J (301) 415-3017 | |
| Shared Package | |
| ML051040354 | List: |
| References | |
| TAC L52650 | |
| Download: ML051040346 (4) | |
Text
OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS SAFETY EVALUATION REPORT RELATED TO A REQUEST TO REVISE AUTHORITY TO DISPOSE OF CONTAMINATED DEMOLITION DEBRIS PURSUANT TO 10 CFR 20.2002 YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION DOCKET NO.50-029
1.0 BACKGROUND
By letter dated December 22, 2004, as supplemented February 7, 2005, Yankee Atomic Electric Company (YAEC, the licensee) requested U. S. Nuclear Regulatory Commission (NRC) approval of proposed procedures for disposal of decommissioning waste at the Waste Control Specialists (WCS) facility in Andrews, Texas. YAEC is requesting to dispose of 27.3 million kilograms (60 million pounds) of demolition debris, which consists of a mixture of steel, soil, asphalt, and concrete. This debris will potentially contain low amounts of residual radioactivity.
To dispose at WCS, which does not have a license to handle or dispose of licensable Atomic Energy Act materials, YAEC is required to demonstrate that the alternate disposal is safe through a 10 CFR 20.2002 request.
The WCS site is a RCRA Subtitle C hazardous waste disposal facility permitted by the State of Texas. Its permit allows it to accept and dispose of certain radioactive materials that are exempt from licensing under the Atomic Energy Act. The site does have requirements in its permit for waste acceptance, environmental monitoring, and radiation protection of workers.
The local climate is arid with approximately 35.5 cm/y (14 inches/y) of precipitation. The site will have a 5-meter (16.5-ft) thick engineered cover over its cells after closure and has a 300-meter (1000-ft) thick unsaturated zone between the base of the disposal cells and the aquifer. The proposed waste volume to be sent to WCS is small compared to the annual volume of all wastes received by WCS and the YAEC waste will not be segregated but rather co-mingled with other waste as it is received.
2.0 TECHNICAL EVALUATION
2.1 SOURCE TERM The licensee has provided characterization of the material to be disposed of based on activity fractions from current surveys. YAEC provided a list of typical radionuclides in the waste stream, isotopic percentage for steel, concrete and soil, and expected surface contamination levels for steel and concrete. H-3 concentrations were provided for RSS concrete and other concrete. These fractions were used to determine the appropriateness of the concentration limits the licensee is requesting for cobalt-60 and cesium-137. The staff determined the
Page radionuclide suite is appropriate for the site. The licensee has also used conservative assumptions of chemical form in its dose analysis.
2.2 SCENARIOS AND PATHWAYS Four general scenarios need to be assessed to investigate the potential doses from the re-use of this material: (1) transportation of material, (2) public worker dose at WCS, (3) potential intruder post-closure issues, and (4) potential effects on groundwater. For this application, the licensee included analysis of three scenarios: various workers involved in transporting the material by truck or rail, worker dose at WCS, and a resident farmer living on top of the disposal site after closure and using the groundwater for all the standard exposure pathways except aquatic foods. To demonstrate that it will meet a few millirem per year criteria, YAEC calculated worker external doses for concrete rubble, scrap iron and soil with residual moisture content for unit concentrations (1 pCi/g) of Co-60 & Cs-137 using transportation and disposal facility scenarios. YAEC identified the maximally exposed worker as a landfill driver receiving 0.62 millirem/yr.
WCS maintains a Radiation Protection Program including routine performance of radiation, contamination, and airborne radioactive material surveys. The facility currently conducts disposal activities involving naturally occurring radioactive materials in concentrations greater than the primary isotopes of concern and the naturally occurring radioactive materials have potential for higher doses from similar concentrations due to inhalation. Despite this much larger internal dose hazard, the site has had no significant internal dose exposures. Therefore, operating experience indicates that there would be no internal dose hazards associated with the proposed disposal.
The staff finds these scenarios to be adequate and reasonable for the assessments required for compliance with 10 CFR 20.2002. Considering the depth of the engineered cover, it is not reasonable to assume that waste will be exposed by intruders onto the site.
2.3 COMPUTER MODELS The licensee used Microshield to calculate the doses to drivers and workers, and used RESRAD version 6.22 in a deterministic mode to calculate the dose for the post-closure resident farmer.
The use of these computer codes is appropriate for the scenarios being modeled. The staff finds these computer codes to be adequate and reasonable for the assessments required for compliance with 10 CFR 20.2002.
2.4 PARAMETER SELECTION The licensee primarily used median values from distributions and site-specific input parameters, with a mix of default parameters. Site-specific information was used, if available. The parameters selected are appropriate for the scenarios being modeled. The staff finds that the controlling parameters in the critical scenario were adequately conservative to use a deterministic analysis.
2.5 SENSITIVITY/UNCERTAINTY CALCULATIONS
Page No sensitivity or uncertainty calculations were performed by the licensee. These are not needed if the deterministic analyses are reasonably bounding as they are in this case.
2.6 LICENSEE RESULTS The licensees results show that the dose is controlled by the external dose to the workers -
specifically the landfill driver or operator. The licensee assumed that only two workers at the facility would be exposed to the entire shipment and that each would be near the waste for approximately 468 hours0.00542 days <br />0.13 hours <br />7.738095e-4 weeks <br />1.78074e-4 months <br /> per year. The staff finds this to be an adequately conservative assumption. Based on these controlling doses, the licensee has requested approval of a maximum concentration of 740 Bq/kg (20 pCi/g) Co-60 and 3700 Bq/kg (100 pCi/g) Cs-137, using a sum of fractions approach. These concentrations are based on a 0.05 mSv/y (5 mrem/y) maximum dose to a worker and is in the upper range of the NRC policy of approving doses of a few mrem for 10 CFR 20.2002 requests. The volumetric or concentration limits will be implemented by the Yankee truck monitoring system.
2.7 INDEPENDENT ANALYSES AND DISCUSSION Along with reviewing the dose analyses provided by YAEC, the staff performed an independent analyses of the potential impact from the waste. The staff used NUREG-1640, Radiological Assessment for Clearance of Materials From Nuclear Facilities, to estimate worker radiation doses. NUREG-1640 provides effective dose equivalent (EDE) per unit mass activity (µSv/y per Bq/g) and surface contamination unit activity (µSv/y per Bq/cm2), and identifies critical groups for steel and concrete processing scenarios from nuclear facility to an end use. The end use scenarios selected for this situation were truck transportation and industrial landfill disposal, and no beneficial re-use was assumed. The staff performed a comparison using the generic analyses for concrete and steel in NUREG-1640 to estimate the overall dose from the waste disposal. For this proposal, the staff used the dose coefficients in Appendix F and I.
The licensee is basing the concentration limits on exposure to the workers at the site. The staff agrees that potential doses from leachate at the site is extremely low, given the natural characteristics of the site. Therefore, the staff agrees that the critical scenario is the landfill worker.
Surface contamination levels are necessary for estimating workers doses, and YAEC provided expected contamination levels for steel and concrete. Worker doses occur principally from external exposure from Co-60 and Cs-137 on or in materials. Typically, the dose factors are orders of magnitude larger for Co-60 and Cs-137 than for other fission or activation products.
Consequently, Co-60 and Cs-137 contribute most of the worker dose, since they are a significant fraction of radionuclides identified. Some low levels of transuranic materials are present on steel, but have minor dose contributions.
Generally, two analyses are performed as part of a 20.2002 request: (1) the maximum concentration in a single shipment and (2) the average dose from the entire disposal. These doses should be to the average member of the critical group. The licensees concentration limits are based on a maximally exposed individual and, therefore, provide a conservative limit for single shipments. The licensee provided limited average concentration information in its submittal. The staff used this information and previous experience to estimate the average
Page dose over the entire disposal. The staff determined that the overall doses to the average member of the critical group will likely be a fraction of a mrem per year.
YAEC plans to use a truck monitoring system to control the activity concentration for shipments.
The licensees report on the calibration and use of the Yankee truck monitoring system was reviewed previously by NRC and determined to be technically acceptable.
3.0 CONCLUSION
S The staffs review finds the licensees proposal to be adequate and reasonable to demonstrate that the dose will be below the NRC policy limit of a few mrem. The licensee used appropriate scenarios and computer models to bound the exposure at their proposed concentration limits for single shipments. In addition, the staff determined that the overall doses to the average member of the critical group will likely be a fraction of a mrem per year. Therefore, we have determined that the proposed disposal of demolition debris from the YNPS at the WCS facility in Andrews, Texas, is acceptable.
Further, in accordance with 10 CFR 30.11, "the Commission may, upon application by an interested person or upon its own initiative, grant such exemptions from the requirements of the regulations...as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest." Based on the above analyses, this material authorized for disposal poses no danger to public health and safety, does not involve information or activities that could potentially impact the common defense and security of the United States, and it is in the public interest to dispose of wastes in a controlled environment such as that provided by the licensed, state-regulated landfills. Therefore, to the extent that this material authorized for disposal in this 20.2002 authorization is otherwise licensable, the staff concludes that the material is exempt from further Atomic Energy Act and NRC licensing requirements.
Docket No.: 050-029 License No.: DPR-003 Principal Contributors: C. McKenney T. Youngblood
Contact:
John B. Hickman, NMSS/DWMEP 301-415-3017