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Notice

Draft Decommissioning Questions and Answers Regarding Clarification of License Termination Guidance of the Nuclear Regulatory Commission's Office Nuclear Material Safety and Safeguards; Notice of Availability

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AGENCY:

Nuclear Regulatory Commission.

ACTION:

Notice of availability and request for public comment.

SUMMARY:

The Nuclear Regulatory Commission's (NRC) Office of Nuclear Material Safety and Safeguards (NMSS) is announcing the availability of draft decommissioning questions and answers regarding clarification of license termination guidance, for public comment.

The Nuclear Energy Institute (NEI) and NRC staff identified an approach to clarify existing guidance associated with the License Termination Rule (10 CFR part 20, subpart E), in concert with NMSS” decommissioning guidance consolidation project. Under this approach, NEI's License Termination Task Force (Task Force) generated questions (Qs) associated with decommissioning issues that are common to the industry. The Task Force also proposed answers (As) to the questions and submitted the Q&As to NRC staff for review. NRC staff reviewed the Q&As and the supporting technical bases and provided comments to NEI on September 28, 2001. An open meeting was held between NRC, NEI, and industry representatives on December 4, 2001, to discuss each Q&A and the technical issues to ensure that the questions were properly asked and answered and were supported by a defensible technical basis. NRC staff and NEI further developed the Q&As so that they adequately reflect NRC regulations and guidance and include a sound technical basis.

As a result of this cooperation, eight Q&As have been found acceptable by NRC staff. Seven of the Q&As were to be incorporated into the draft document “Consolidated NMSS Decommissioning Guidance: Characterization, Survey, and Determination of Radiological Criteria” (NUREG-1757, Volume 2) to solicit public comment on them. However, two Q&As were inadvertently omitted. Therefore, five Q&As are included in Volume 2 of NUREG-1757, and three Q&As are included in the “supplementary information” section of this notice. Volume 2 of NUREG-1757 is being published for public comment on or close to the date of this notice. NRC is seeking public comment on the Q&As and Volume 2 of NUREG-1757 in order to receive feedback from the widest range of interested parties and to ensure that all information relevant to developing the document is available to the NRC staff. These draft documents are being issued for comment only and are not intended for interim use. The NRC will review public comments received on the draft documents. Suggested changes will be incorporated, where appropriate, in response to those comments, and a final document will be issued for use. The final Q&As will be included in the text of the final document of Volume 2 of NUREG-1757.

DATES:

Comments on this draft document should be submitted by December 26, 2002. Comments received after that date will be considered to the extent practicable.

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ADDRESSES:

Members of the public are invited and encouraged to submit written comments to: Duane W. Schmidt, Project Manager, Office of Nuclear Material Safety and Safeguards, Mail Stop T-7F27, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Hand-deliver comments to: 11555 Rockville Pike, Rockville, MD, between 7:30 a.m. and 4:15 p.m., Federal workdays. Comments may also be sent electronically to decomcomments@nrc.gov. Copies of comments received may be examined at the ADAMS Electronic Reading Room on the NRC web site, and the NRC Public Document Room, 11555 Rockville Pike, Room O-1F21, Rockville, MD 20852. The NRC Public Document Room is open from 7:45 a.m. to 4:15 p.m., Monday through Friday, except on Federal holidays.

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FOR FURTHER INFORMATION, CONTACT:

Duane W. Schmidt, Mail Stop T-7F27, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Telephone: (301) 415-6919; Internet: dws2@nrc.gov.

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SUPPLEMENTARY INFORMATION:

Qeustion 5

What are acceptable methods to characterize embedded piping and buried piping?

Answer to Question 5

Several methods have been used to characterize the residual activity within embedded pipe, and these methods can be used for buried piping, as well. By definition, “embedded piping” is piping (e.g., part of a plant system) that is found in buildings and encased in concrete floors and walls, while “buried piping” is piping (e.g., culvert) that is buried in soils. To be found acceptable, the methods must each address the following issues:

  • Radionuclides of interest and chosen surrogate
  • Levels and distribution of contamination
  • Internal surface condition of the piping
  • Internal residues and sediments and their radiation attenuation properties
  • Removable and fixed surface contamination
  • Instrument sensitivity and related scan and fixed minimum detectable concentrations
  • Piping geometry and presence of internally inaccessible areas/sections
  • Instrument calibration
  • Data quality objectives (DQOs)

An industry study (Cline, J. E., “Embedded Pipe Dose Calculation Method,” Electric Power Research Institute Report No. 1000951, November, 2000) evaluated several techniques for measuring the radiological contamination on the inside of embedded pipe. Measurement techniques included pipe crawlers, gamma-ray scanners, dose rate measurements with dose-to-curie computations, scraping samples with radiochemical analyses, and smear samples with radiochemical analyses. A brief description of these methods is provided below.

The pipe crawler uses a beta sensitive detection system that is inserted into the pipe with a cable. Spacers keep the detectors at a fixed distance from the pipe wall. Measurements can be made at various points or as a continuous scan within the pipe to provide a profile of the extent and distribution of the contamination. Scaling factors based on a laboratory radiochemistry analysis of the deposited material can be applied to the measurements to provide radionuclide quantities in the pipe.

The gamma-ray scanner uses a calibrated, collimated high-purity germanium or sodium iodide spectrometer to make external measurements on the pipe. This gamma-ray scanning yields an average concentration over the length of the pipe within the field of view of the detector. The sensitivity of this method may be limited by the thickness of the piping itself and concrete between the pipe and the detector. Some radionuclide identification is possible and scaling factors can be applied as discussed above for the pipe crawler.

The dose rate measurements are also made on the external surface of the walls or floors containing the embedded pipe using a sensitive gamma detector capable of reading in the roentgen per hour range. The dose rate readings may be used directly in determining compliance with the dose criteria or used to make dose-to-curie conversions based on other measurements providing radionuclide identification.

Radionuclide identification for the contamination in the pipe may be accomplished by smear or scraping samples and radiochemical analysis. The industry report compared radionuclide ratios determined by smears and by scrapings with those found by etching the surface of the pipe. The report concluded that either of these techniques yields radionuclide mixes that are representative of the average total deposits.

Each approach is useful in specific applications, and multiple methods might be used in complex facilities like power plants. Each method also has limitations and uncertainties that must be addressed.

Other useful information on embedded pipe characterization may be found in sources, such as the U.S. Department of Energy Innovative Technology Reports and case studies published in open literature.

Regardless of the source of the information, it is incumbent on the licensee to develop and document a comprehensive approach to embedded pipe and buried piping characterization that accounts for limitations and uncertainties, taking into account the Multi-Agency Radiation Survey and Site Investigation Manual (NUREG-1575, Rev. 1) guidance in developing the related DQOs. It should also specifically address each of the critical issues in the bulleted list above.

Question 9

Is the collection of additional characterization data, beyond that available from periodic radiation protection surveys, required in the license termination plan for structures, components, and soils that will be removed from the facility prior to license termination?

Answer to Question 9

No. In general, radiological data obtained during characterization surveys are used to determine the radiological status of the site, including facilities, buildings, surface and subsurface soils, and surface and ground water. In turn, this information is used to support the planning and design of final status surveys (FSS). In addition to providing the basis of the design of FSS, characterization surveys are used to support the following:

  • Identification of remaining site dismantlement activities
  • Development of new (or revisions to existing) remediation plans and procedures
  • Revisions to decommissioning costs and trust fund
  • Identification of environmental aspects not previously considered
  • Revisions to the Environmental Report

Since the license termination process is only concerned with the status of facilities after the completion of all remediation activities, radioactivity associated with structures, components, and soils that will be removed from the facility and appropriately disposed of elsewhere, is not an issue as it cannot contribute to public dose controlled under 10 CFR 20.1402—“Radiological Criteria for Unrestricted Use.” Therefore, additional characterization data need not be collected. Start Printed Page 60706

Question 10

Is characterization data required to support initial classification of Class 1 areas?

Answer to Question 10

Areas classified as Class 1 do not require characterization data to support that classification.

Note:

Characterization data are needed to support decommissioning activities for all areas including:

  • Determination of radionuclide distribution profiles and identification of surrogate radionuclides
  • Dose modeling and development of derived concentration guideline levels
  • Final status survey design and instrument selection
  • Structuring the data quality objectives
  • Assessment of spatial variability of radioactive contaminants on building surfaces and in surface and subsurface soils
  • Assessment of whether ground water is impacted, using the results of the surface and subsurface soil characterization surveys
  • Initially defining and changing the boundaries of Class 1 survey units with bordering and adjacent survey units
  • Re-classification of survey units (using guidance in NUREG-1575, “Multi-Agency Radiation Survey and Site Investigation Manual,” and NUREG-1727, “NMSS Decommissioning Standard Review Plan”)
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Dated at Rockville, MD, this 20th day of September, 2002.

For the Nuclear Regulatory Commission.

Larry W. Camper,

Chief, Decommissioning Branch, Division of Waste Management, Office of Nuclear Material Safety and Safeguards.

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[FR Doc. 02-24442 Filed 9-25-02; 8:45 am]

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