ML061220260

Letter to Steven L. Ceccio from Patrick Isaac University of Michigan Ford Nuclear Reactor - Amendment Decommissioning Plan Approval

ML061220260
Person / Time
Site:	University of Michigan
Issue date:	06/26/2006
From:	Isaac P NRC/NRR/ADRA/DPR/PRTA
To:	Ceccio S University of Michigan
Issac P, NRR/DRIP/REXB, 415-1019
References
TAC MC3707
Download: ML061220260 (88)

Text

June 26, 2006 Mr. Steven L. Ceccio, Director Phoenix Memorial Laboratory 2301 Bonisteel Boulevard University of Michigan Ann Arbor, MI 48109

SUBJECT:

UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTORAMENDMENT RE: DECOMMISSIONING PLAN APPROVAL (TAC NO. MC3707)

Dear Mr. Ceccio:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 50 to Facility Operating License No. R-28 for the University of Michigan Ford Nuclear Reactor, Docket No. 50-2.

The amendment approves the decommissioning plan (DP) for the Ford Nuclear Reactor in response to your application of June 18, 2004, as supplemented on June 23, 2004, January 5, and January 10, 2006. The amendment authorizes the inclusion of the approved DP as a supplement to the safety analysis report pursuant to Title 10, Section 50.82(b)(5), of the Code of Federal Regulations (10 CFR 50.82(b)(5)). In addition, in accordance with 10 CFR 50.82(b)(5), the NRC staff has added license conditions to Facility Operating License No. R-28 deemed appropriate and necessary for approval of the DP.

We have also enclosed a copy of the safety evaluation supporting Amendment No. 50.

Sincerely,

/RA/

Patrick Isaac, Project Manager Research and Test Reactors Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-02

Enclosures:

1. Amendment No. 50

2. Safety Evaluation cc w/enclosures: See next page

University of Michigan Docket No. 50-02 cc:

Special Assistant to the Governor Office of the Governor Room 1State Capitol Lansing, MI 48909 Mr. C.W. Becker Phoenix Memorial Laboratory 2301 Bonisteel Boulevard University of Michigan Ann Arbor, MI 48109 Michigan Department of Environmental Quality Waste and Hazardous Materials Division Hazardous Waste and Radiological Protection Section Nuclear Facilities Unit 525 West Allegan Street P.O. Box 30241 Lansing, MI 48909-7741 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

June 26, 2006 Mr. Steven L. Ceccio, Director Phoenix Memorial Laboratory 2301 Bonisteel Boulevard University of Michigan Ann Arbor, MI 48109

SUBJECT:

UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTORAMENDMENT RE: DECOMMISSIONING PLAN APPROVAL (TAC NO. MC3707)

Dear Mr. Ceccio:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 50 to Facility Operating License No. R-28 for the University of Michigan Ford Nuclear Reactor, Docket No. 50-2.

The amendment approves the decommissioning plan (DP) for the Ford Nuclear Reactor in response to your application of June 18, 2004, as supplemented on June 23, 2004, January 5, and January 10, 2006. The amendment authorizes the inclusion of the approved DP as a supplement to the safety analysis report pursuant to Title 10, Section 50.82(b)(5), of the Code of Federal Regulations (10 CFR 50.82(b)(5)). In addition, in accordance with 10 CFR 50.82(b)(5), the NRC staff has added license conditions to Facility Operating License No. R-28 deemed appropriate and necessary for approval of the DP.

We have also enclosed a copy of the safety evaluation supporting Amendment No. 50.

Sincerely,

/RA/

Patrick Isaac, Project Manager Research and Test Reactors Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-02

Enclosures:

1. Amendment No. 50

2. Safety Evaluation cc w/enclosures: See next page DISTRIBUTION:

PUBLIC PRT r/f JQuichocho PIsaac CBassett EHylton MMendonca AAdams OGC DHArrison TDragoun KWitt DHughes WSchuster MVoth GHill (2) (T5-C3) BThomas ADAMS ACCESSION NO: ML061220260 OFFICE TechEd PRT:LA PRT:RI PRT:PM OGC PRT:BC NAME PIsaac for EHylton:tls* PIsaac for TDragoun* PIsaac* HWedewer* BThomas:tls*

DATE 5/11/06 5/15/06 5/15/06 5/16/06 6/2/06 6/26/06 OFFICIAL RECORD COPY

UNIVERSITY OF MICHIGAN DOCKET NO. 50-02 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. R-28

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application filed by the University of Michigan (the licensee), dated June 18, 2004, and as supplemented on June 23, 2004, January 5, and January 10, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Title 10, Chapter 1, of the Code of Federal Regulations (10 CFR Chapter 1);

B. The facility will be possessed and decommissioned in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E. This amendment is issued in accordance with 10 CFR Part 51 of the regulations of the Commission and all applicable requirements have been satisfied; and

2. Accordingly, the license is amended by changes to the following paragraph which is hereby amended to read as follows:

1.B. The facility will be possessed and decommissioned in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

3. Accordingly, the license is amended by changes to paragraph 2.C.3 to Facility Operating License No. R-28 which hereby reads as follows:

2.C.3. Decommissioning

a. The license is amended to approve the decommissioning plan described in the

licensees application dated June 23, 2004, as supplemented on January 5, 2006, and January 10, 2006, and authorizes inclusion of the decommissioning plan as a supplement to the safety analysis report pursuant to 10 CFR 50.82(b)(5).

b. The licensee may make changes to the decommissioning plan without prior approval provided the proposed changes do not:

(i) Require Commission approval pursuant to 10 CFR 50.59; (ii) Use a statistical test other than the Sign test or Wilcoxon Rank Sum test for evaluation of the final status survey; (iii) Increase the radioactivity level, relative to the applicable derived concentration guideline level, at which an investigation occurs; (iv) Reduce the coverage requirements for scan measurements; (v) Decrease an area classification (i.e., impacted to unimpacted; Class 1 to Class 2; Class 2 to Class 3; or Class 1 to Class 3);

(vi) Increase the Type I decision error; (vii) Result in more than a minimal increase in the environment consequences not previously evaluated in the final safety analysis report (as updated);

(viii) Foreclose the release of the site for possible unrestricted use.

c. The licensee shall submit reports of all characterization surveys performed that were not part of the license amendment application and shall submit the completed final status survey plan for review prior to performing the final status survey.

4. This license amendment is effective as of the date of its issuance.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION

/RA/

Brian Thomas, Branch Chief Research and Test Reactors Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Date of Issuance: June 26, 2006

SAFETY EVALUATION RELATED TO THE DECOMMISSIONING OF THE UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTOR UNIVERSITY OF MICHIGAN June 2006 Office of Nuclear Reactor Regulation Division of Regulatory Improvement Programs Operating Reactor Improvements Program

ABSTRACT This safety evaluation summarizes the findings of a technical review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation. The staff conducted this review in response to an application filed by the University of Michigan (UM or the licensee) for approval of the decommissioning plan (DP) for the Ford Nuclear Reactor (FNR). The FNR is located on the UM campus in Ann Arbor, Michigan. On the basis of this review, the staff concludes that UM can safely dismantle the FNR and dispose of the component parts in accordance with their DP, as amended, and the NRCs rules and regulations.

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CONTENTS Page ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2.0 BACKGROUND

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.1 Regulatory Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.2 Site and Facility Description and Operating History . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.3 Scope of the Decommissioning Project . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.1 Decommissioning Alternative . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.1.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.2 Controls and Limits on Procedures and Equipment to Protect Occupational and Public Health and Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.2.1 Project Management Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.2.1.1 Decommissioning Organization and Responsibilities . . . . . . . . . . . . 12 3.2.1.2 Key Licensee Positions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.2.1.3 Decommissioning Prime Contractor . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.2.1.4 Safety Review Committee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.2.1.5 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.2.2 Occupational and Public Health and Safety . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.2.2.1 Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 3.2.2.2 Health Physics Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.2.3 Control of Radioactive Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 3.2.2.4 Dose Estimates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.2.2.5 Radioactive Waste Management . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 3.2.3 Training Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.2.3.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.2.4 General Industrial Safety Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.2.4.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.2.5 Radiological Accident Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.2.5.1 Fire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.2.5.2 Pool Leak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 3.2.5.3 Tritium-Loaded Heavy-Water Spill . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 3.2.5.4 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 3.3 Decommissioning Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 3.3.1 Radiological Status of the Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 3.3.1.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 3.3.1.2 Principal Radioactive Components . . . . . . . . . . . . . . . . . . . . . . . . . . 40 3.3.1.3 Sanitary Sewer Lines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 3.3.1.4 Soil beneath the Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 3.3.1.5 Ground Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 3.3.1.6 Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 3.3.1.7 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47

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CONTENTS (Continued)

Page 3.3.2 Radiological Release Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 3.3.2.1 Structure Surfaces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 3.3.2.2 Surface Soil and Sediment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 3.3.2.3 Subsurface and Inaccessible Structures . . . . . . . . . . . . . . . . . . . . . . 52 3.3.2.4 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 3.3.3 Decommissioning Tasks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 3.3.3.1 Characterization Surveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 3.3.3.2 Dismantlement and Decontamination of the Facility . . . . . . . . . . . . . 53 3.3.3.3 Final Survey and Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 3.3.3.4 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 3.3.4 Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 3.3.4.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 3.3.5 Proposed Final Status Survey Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 3.3.5.1 General Survey Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 3.3.5.2 Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 3.3.5.3 Data Quality Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 3.3.5.4 Classifications of Areas by Contamination Potential . . . . . . . . . . . . . 60 3.3.5.5 Identification of Survey Units . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 3.3.5.6 Demonstrating Compliance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 3.3.5.7 Background Reference Areas and Materials . . . . . . . . . . . . . . . . . . . 62 3.3.5.8 Final Status Survey Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 3.3.5.9 Data Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 3.3.5.10 Final Status Survey Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 3.3.5.11 Change Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 3.3.5.12 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 3.4 Estimated Cost . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 3.4.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 3.5 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 3.5.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 3.5.2 Quality Assurance for Design, Construction, Testing, Modification, and Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 3.5.3 Quality Assurance for Packaging, Preparation for Shipment, and Transportation of Licensed Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 3.5.4 Quality Assurance for Final Status Survey and Associated Documentation . . 68 3.5.4.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 3.5.4.2 Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 3.5.4.3 Written Quality Assurance Program . . . . . . . . . . . . . . . . . . . . . . . . . 69 3.5.4.4 Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 3.5.4.5 Quality Assurance Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 3.5.4.6 Control of Measuring Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 3.5.4.7 Audits and Corrective Actions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 3.5.4.8 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 3.6 Physical Security . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 3.6.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71

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CONTENTS (Continued)

Page 4.0 ADDITIONAL LICENSE CONDITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 4.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 5.0 TECHNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 5.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73

6.0 ENVIRONMENTAL CONSIDERATION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73

7.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77 LIST OF FIGURES Figure 2-1 UM FNR Site Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Figure 2-2 FNR Basement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Figure 2-3 FNR First Floor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Figure 2-4 FNR Second Floor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 Figure 2-5 FNR Third Floor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Figure 2-6 FNR Fourth Floor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Figure 2-7 East-West Cross Section of Reactor Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 Figure 3-1 Organization Chart for the FNR Decommissioning Project . . . . . . . . . . . . . . . . . . 14 Figure 3-2 Radiation Levels (R/hr) on the Reactor Grid Plate (April 2004) . . . . . . . . . . . . . . 42 LIST OF TABLES Table 2-1 Profile of UM FNR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Table 3-1 Quantities of Individual Expected Radionuclides Producing the Emission of the AEC during an 8-Hour Fire . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 Table 3-2 Radioactivity of the Reactor Pool Water (March 17, 2004) . . . . . . . . . . . . . . . . . . 41 Table 3-3 Estimated Material Volumes for the Thermal Column . . . . . . . . . . . . . . . . . . . . . . 44 Table 3-4 List of Potential Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 Table 3-5 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 Table 3-6 Acceptable License Termination Screening Values of Common Radionuclides for Surface Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 Table 3-7 Instrumentation for FNR Radiological Surveys . . . . . . . . . . . . . . . . . . . . . . . . . . . 59

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1.0 INTRODUCTION

By letter dated June 18, 2004, the University of Michigan (UM or the licensee) (Ref. 1) submitted a license amendment request to the U.S. Nuclear Regulatory Commission (NRC) for approval of its decommissioning plan (DP), Revision 00, dated June 23, 2004 (Ref. 2), and authorization to dismantle and dispose of component parts of the Ford Nuclear Reactor (FNR).

Subsequently, the licensee submitted Revision 1 to the DP (Ref. 3), dated January 5, 2006.

On January 10, 2006 (Ref. 4), the licensee submitted additional detail concerning site characterization.

The licensee selected the DECON option as the decommissioning alternative. This option will consist of decontamination and removal of equipment and material containing residual radioactivity from the site to levels allowing for unrestricted release as specified in Title 10, Section 20.1402 of the Code of Federal Regulations (10 CFR 20.1402). In the license amendment request, UM described its plan for developing and implementing the final status survey (FSS) plan to verify and document that the decommissioned areas and structures meet the requirements of release for unrestricted use. Upon completion of decommissioning-related activities and the FSS, UM will submit the necessary documentation for review and approval by the NRC to support license termination.

The NRC published a Notice and Solicitation of Comments Pursuant to 10 CFR 20.1405 and 10 CFR 50.82(b)(5) Concerning Proposed Action to Decommission the University of Michigan Ford Nuclear Reactor (FNR) in the Federal Register on September 8, 2004 (69 FR 54326-54327), and in The Ann Arbor News on September 9, 2004. The agency did not receive any comments.

2.0 BACKGROUND

2.1 Regulatory Basis The contents of proposed DPs for research and test reactors must include the following, as specified in 10 CFR 50.82(b)(4):

• the choice of the alternative for decommissioning with a description of activities (see Section 3.1 below)

• a description of the controls and limits on procedures and equipment to protect occupational and public health and safety (see Section 3.2 below)

• a description of the planned FSS (see Section 3.3.5 below)

• an updated cost estimate for the chosen alternative for decommissioning, comparison of that estimate with present decommissioning funds set aside, and plan for assuring the availability of adequate funds to complete decommissioning (see Section 3.4 below)

• a description of quality assurance provisions, physical security plan provisions, and technical specifications (TSs) in place during decommissioning (see Sections 3.5, 3.6, and 5.0 below)

According to 10 CFR 50.82(b)(5), if the DP demonstrates that the decommissioning will be performed in accordance with the regulations in this chapter and will not be inimical to the common defense and security or to the health and safety of the public, and after notice to interested persons, the Commission will approve, by amendment, the plan subject to such conditions and limitations as it deems appropriate and necessary. The agency based license conditions for this amendment on Appendix 2 to NUREG-1700, Revision 1, Standard Review Plan for Evaluating Nuclear Power License Termination Plans (Ref. 5). Furthermore, the staff established a license condition in accordance with the requirement of 10 CFR 50.82(b)(5) stating that the approved DP will be a supplement to the safety analysis report (SAR) or equivalent.

As specified in 10 CFR 50.82(b)(6), the Commission will terminate the license if it determines that the decommissioning was in accordance with the approved DP, and that the FSS and associated documentation demonstrate that the facility and site are suitable for unrestricted release in accordance with criteria for decommissioning in 10 CFR Part 20, Standards for Protection Against Radiation, Subpart E, Radiological Criteria for License Termination.

2.2 Site and Facility Description and Operating History The FNR site and facility is situated on property owned by UM in Ann Arbor, Michigan. The FNR is located on the North Campus of UM, which is a tract of 900 acres, approximately 1.25 miles northeast of the central business district of Ann Arbor.

The reactor building is a windowless, four-story, reinforced concrete building supported by an integral post-and-beam structure. The 12-inch exterior walls of the reactor building are integral with the footings and foundation mats. The building is approximately 70 feet wide, 68 feet long, and 69 feet high with 55 feet and 46 feet of the structure exposed above grade on the east and west, respectively (Ref. 2).

In 1954, the Phoenix Memorial Laboratory (PML) was completed, and construction of the FNR began in 1956. The integrated power generated during operation of the FNR was estimated at 17,868 megawatt-days between 1957 and 2003 (Ref. 2). The FNR was an open-pool research reactor of the materials testing reactor design. It was a light-water moderated and cooled nonpower reactor with a heterogeneous core composed of aluminum and enriched uranium-235. The reactor was licensed to operate at 2 megawatt thermal (Mwt) power. After the initial startup of the FNR in 1957, reactor operations were permanently placed in safe-shutdown mode on July 3, 2003.

Although conjoined with the FNR, PML is not part of the DP. PML is a four-story, reinforced concrete building supported by an integral post-and-beam structure that contains offices, wet and dry laboratories, a machine shop, two hot cells, a cobalt irradiator, and various equipment and storage rooms.

To assist the licensees plans to decommission the FNR, the NRC amended License No. R-28 on January 29, 2004 (Amendment No. 47), and on May 1, 2006 (Amendment No. 48), to support cessation of reactor operations. Prior to amending the license, the NRC required the licensee to remove all reactor fuel elements from the FNR and return these licensed materials to the U.S. Department of Energy (DOE).

Table 2-1 Profile of UM FNR General Reactor Information:

Owner: UM Operator: UM Licensee: UM Architect/Engineer: Smith, Hinchman & Grylls, Inc.

Nuclear Design: Babcock & Wilcox Co.

Construction: Jeffress-Dyer, Inc. and Babcock & Wilcox Co.

Principal Uses: Training and research Reactor Operation and Authorization:

Initial Criticality: September 19, 1957, 04:00 Date Secured: July 3, 2003, 15:37 NRC Utilization Facility License #: R-28 NRC Facility Docket #: 50-2 Maximum Power, Steady State, Mwt: 2 f thermal Steady State, Water Reflected (nv): 3 x 1013 n cm-2 s-1 peak Specific Power (kW/kg235U): 382.4 Core Power Density, (kW/l): 8.5 Fuel Material: UAIx, U3O8 Uranium Enrichment, % 235U: <20%

Fuel Element Geometry: MTR18 fuel plates (3.25 in. x 2.94 in. x 34.78 in.)

Element Cladding Material: Aluminum Element Cladding Thickness: 0.06 in.

Core Configuration: 35-40 MTR plate-type fuel elements Core Active Height: 24.0 in.

No. of Available Fuel Positions: 48 Coolant: Light Water Moderator: Light Water Reflector: Light water with heavy water on the north face The following systems continue in operation:

• FNR building utility services that are required for facility surveillance and maintenance under possession-only status

• FNR manually actuated and automated fire alarm systems

• FNR security and radiological alarm systems

• FNR water demineralization system 2.3 Scope of the Decommissioning Project The DP lists the various areas, structures, and components that are included in the decommissioning project. Some of the specific areas include the reactor pool and associated structures and systems, pneumatic tube system, cooling system, storage ports, building crane, foundation tile, soil under and around the reactor pool, and other impacted interior and exterior building surfaces (see Figures 2-1 to 2-7). The FSS will include the entire FNR facility, such as the building, systems, and any other areas, as necessary. Residual radioactivity present in these structures and components will be decontaminated and/or decommissioned to levels that will allow for the unrestricted use of this site.

Figure 2-1 UM FNR Site Plan Figure 2-2 FNR Basement Figure 2-3 FNR First Floor Figure 2-4 FNR Second Floor Figure 2-5 FNR Third Floor Figure 2-6 FNR Fourth Floor Figure 2-7 East-West Cross Section of Reactor Pool 3.0 EVALUATION The NRC staff has reviewed the licensees proposed actions to decontaminate, dismantle, and dispose of component parts of the FNR, and to perform an FSS. In addition, the staffs review focused on the licensee meeting the regulatory requirements discussed in Section 2.1 above and included consideration of the following:

• management responsibilities/commitments and personnel qualifications to continue following applicable regulations, regulatory guides, standards, and health and safety plans, including procedures

• use of appropriate equipment and instrumentation, radiation survey methods, training, personnel dosimetry, and radioactive waste disposal

• the plan to develop and perform the FSS of the facility

• the commitments needed to implement an adequate quality assurance plan

• the methods that the licensee will use to meet the radiological release criteria 3.1 Decommissioning Alternative The licensees stated objective of decommissioning the FNR is the release of the site for unrestricted use. As such, the licensee selected DECON as the preferred decommissioning alternative needed to accomplish the stated objective.

The licensee will decontaminate facility equipment and structural components to minimize radioactive waste. Structural portions of the building and materials found to be radiologically contaminated and/or activated will be decontaminated, sectioned and removed, and/or processed, as necessary. These activities will be followed by an extensive and comprehensive FSS to demonstrate compliance with cleanup criteria, and thus allow for release of the site for unrestricted use. To support license termination, the licensee will document the results of this FSS in a report to be submitted to the NRC for review and approval.

3.1.1 Conclusions The NRC staff has concluded that the choice of DECON and associated proposed plans meet the provisions of 10 CFR 50.82(b)(4)(i) for decommissioning without significant delay and are, therefore, acceptable.

3.2 Controls and Limits on Procedures and Equipment to Protect Occupational and Public Health and Safety 3.2.1 Project Management Structure 3.2.1.1 Decommissioning Organization and Responsibilities The licensee will continue to retain ultimate responsibility for full compliance with the existing NRC reactor license and the applicable regulatory requirements during decommissioning.

The responsibility for the decommissioning is assigned to the Executive Vice President and Chief Financial Officer. The Executive Vice President has established, through the Associate Vice President for Facilities and Operation, a project organization to oversee the decommissioning of the FNR as shown in Figure 3-1. The Director of Occupational Safety and Environmental Health leads the FNR project staff and is responsible for the facilitys license and authorizing the expenditure of funds on decommissioning activities. The reactor manager remains responsible for ensuring that decommissioning-related activities are conducted in a safe manner within the limitations of the facilitys license and in compliance with applicable Federal, State, and local regulations. The radiation safety officer (RSO), who is organizationally independent of the reactor manager, remains responsible for radiological safety at the facility.

A safety review committee, chaired by a representative of the Vice President for Research, is responsible for overseeing decommissioning activities to ensure they are performed safely and in accordance with all applicable license requirements and Federal, State, and local regulations.

Figure 3-1 Organization Chart for the FNR Decommissioning Project Regents University of Michigan President Vice President Executive Vice President Research Chief Financial Officer Associate Vice President Facilities & Operations Director Chair, Review Committee Occupational Safety and Environmental Health Review Committee Radiation Safety Officer Reactor Manager Safety Reactor Staff Staff Technical, Safety & Operational, Environmental Quality &

Management Licensing Management Prime Contractor Project Manager Health Physics Supervisor Prime Contractor Staff Subcontractors, Testing Laboratories, Vendors, Shippers, etc.

3.2.1.2 Key Licensee Positions The licensee will maintain the key management positions described below to support the decommissioning of the FNR.

The Director of Occupational Safety and Environmental Health (Director) has oversight authority and is responsible for the following:

• the facilitys license (compliance and amendments)

• successful completion of decommissioning activities

• authorizing the expenditure of funds for decommissioning

• requesting termination of the license for the FNR

• approval of contractors, subcontractors, and consultants

• approval of budgets and schedules

• serving as technical spokesman for UM on decommissioning activities

• ensuring that the conduct of decommissioning complies with all applicable licenses and registrations held by UM and with compliance to applicable Federal, State, and local regulatory requirements Section 5.2 of the DP lists proposed changes to the TSs to update the qualifications for this position.

The reactor manager has responsibility for the following:

• controlling and maintaining safety and protection of the environment during decommissioning

• determining facility staffing and organization

• ensuring that decommissioning activities are within budgetary and schedule requirements

• reporting performance to the Director and the safety review committee

• approving changes to the facility that satisfy the equivalent requirements of 10 CFR 50.59, Changes, Tests and Experiments, contained in the license

• providing licensing interface with the NRC, Michigan Department of Environmental Quality, and other regulatory agencies

• providing technical oversight and guidance

• reviewing work procedures, radiation work permits (RWPs), and job hazard analyses (JHAs)

• ensuring that shipments of radioactive/hazardous materials are prepared and transported safely and in accordance with all applicable regulations and requirements of the receiver

• acting as interface between contractor, subcontractors, or consultants and the Director or safety review committee

• coordinating staff, contractor, subcontractor, or consultant activities

• providing technical support to the Director and safety review committee

• ensuring that all staff, contractors, and other UM staff supporting decommissioning effectively implement all quality assurance program(s) requirements

• investigating off-normal occurrences or audit findings, scheduling corrective actions, including measures to prevent recurrence of significant conditions adverse to quality, and notifying the Director and each safety review committee member of action taken or planned to be taken

• Assisting the Director in ensuring that decommissioning activities comply with all applicable license requirements and with applicable Federal, State, and local regulations Section 5.2 of the DP lists proposed changes to the TSs to update the qualifications for this position.

The RSO is responsible for the following:

• maintaining the radiation safety and health aspects of programs or procedures and ensuring compliance with programs or procedures

• determining facility radiation safety staffing and organization

• reviewing work procedures, RWPs, and JHAs in situations that could affect potential radiation exposure or safety

• providing technical support to the Director and safety review committee

• ensuring procedures and practices are established to ensure that radiation exposures to the public and facility personnel are kept at as low as reasonably achievable (ALARA) levels

• identifying locations, operations, or conditions that have the potential for significant exposures to radiation or radioactive materials and initiating actions to minimize or eliminate unnecessary exposures

• monitoring contractor and subcontractor health physics coverage of decommissioning activities

• monitoring collective dose for decommissioning activities

• ensuring the implementation of industrial safety, industrial hygiene, and environmental protection programs that comply with all applicable license requirements and with applicable Federal, State, and local regulations Section 5.2 of the DP lists proposed changes to the TSs to update the qualifications for this position.

3.2.1.3 Decommissioning Prime Contractor The licensee provided its criteria for the selection of a prime contractor to manage and supervise all or part of the FNR decommissioning project. The selected prime contractor will manage and supervise operations and services such as characterization, dismantlement, decontamination, waste handling, and quality assurance. UM will select the prime contractor through an evaluation of the following criteria:

• the prime contractors ability to perform the required task as demonstrated by the quality of information provided in a statement of qualification package

• qualifications of key individuals, including but not limited to the key contractor individuals identified in this section, based upon internal and license requirements

• past performance of the contractor and identified key subcontractors with respect to compliance with all Federal, State, and local regulations

• safety record of the contractor and key subcontractors

• relevant experience of contractor and key subcontractors, particularly with decommissioning of research reactors

• references from owners and Federal, State, and local authorities on previous decommissioning projects for which the contractor and key subcontractors participated

• example work products (e.g., RWPs, JHAs, characterization studies, work packages, quality assurance procedures, etc.) provided by the contractor and key subcontractors

• financial qualifications of the contractor and key subcontractors to complete the project The prime contractor will establish and maintain a project manager who will serve as the overall project manager and a vital member of the project team. The prime contractor will also establish and maintain a health physics supervisor to be responsible for providing basic radiation safety support for contractor and subcontractor activities. The prime contractor may retain subcontractors or hire consultants to help in the performance of all or part of the FNR decommissioning project with the prior approval of the Director.

3.2.1.4 Safety Review Committee UM will establish a safety review committee to review decommissioning activities and advise the Director in matters relating to the health and safety of the project.

The safety review committee (as detailed in the DP) will be composed of a chair and a minimum of three members and alternates. The Vice President for Research will appoint the members and alternates. The safety review committee chair will be appointed from the UM faculty, shall have a degree in engineering or a scientific field, and will have a thorough understanding of the decommissioning project. The remaining members of the safety review committee (including alternates) will collectively represent a broad spectrum of expertise appropriate for the decommissioning of the FNR and may be either from within or outside UM.

The safety review committee will meet at least semiannually throughout the duration of decommissioning until completion of the FSS. After completion of the FSS, the safety review committee will meet as necessary to review or approve such matters as desired by the committee chair, the Director, reactor manager or the RSO. The safety review committee will have approval, review, and audit functions as described below.

The safety review committee will approve the following:

• proposed changes in the license or TSs

• proposed changes to the facility that can be implemented without the prior approval of the NRC in accordance with 10 CFR 50.59

• proposed changes in the DP that can be implemented without the prior approval of the NRC

• new procedures and proposed changes to the procedures for the following activities which will be in effect and followed:

normal operation of all systems structures or components described in the TSs or which are important to safety actions for responding to emergency conditions involving the potential or actual release of radioactivity, including provisions for evacuation, reentry, recovery, and medical support actions to be taken to correct off-normal events and specific malfunctions of systems, structures, or components described in the TSs or which are important to safety activities performed to satisfy a surveillance requirement contained in the TSs radiation and radioactive contamination control physical security of the facility implementation of the quality controls for the calibration and response testing of radiation instrumentation used for direct measurement in support of characterization, the FSS, or other quality assurance activities The safety review committee, in its review function, will consider the following:

• regulatory violations and reportable occurrences made pursuant to license and regulatory requirements

• audit reports issued by a member or subcommittee of the safety review committee developed to satisfy any requirement of the committees audit function

• plans for the following decommissioning activities prior to their implementation:

any activity which could compromise the structure and integrity of the reactor pool or the primary coolant system while pool water is relied upon for shielding of irradiated reactor components the dismantlement of the irradiated reactor components in preparation for disposal the movement of any heavy objects greater than 5 tons in weight any activity that could compromise the structural integrity of the post-and-beam structure that supports the reactor building any activity that will result in the direct release of radioactivity from the facility to the sanitary sewer or a navigable waterway the draining of the reactor pool the decontamination or dismantlement of the reactor pool structure any activity for which it is estimated that the cumulative radiation exposure for the activity will exceed 1 person-rem, or an individual radiation exposure to either an occupationally exposed person or a member of the public that could exceed 20 percent of any applicable exposure limits of 10 CFR Part 20 any activity, known or anticipated by the safety review committee, which it requests to review, subject to the approval of the Director The safety review committee, as an audit function, will ensure that the following are independently monitored or audited:

• decommissioning operations to ensure they are performed safely and in accordance with all applicable licenses held by UM and in compliance with applicable Federal, State, and local regulatory requirements

• the quality assurance program to verify that performance criteria are met as well as to determine the effectiveness of the program in satisfying the quality assurance requirements of the decommissioning plan and 10 CFR Part 71, Packaging and Transportation of Radioactive Material 3.2.1.5 Conclusions The licensee has committed to maintaining an adequate organizational structure to oversee and safely manage the decommissioning of the FNR. The staff has determined that the project management structure for the decommissioning of the FNR is consistent with the guidance provided in Appendix 17.1 to NUREG-1537, Revision 0, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, issued February 1996 (Ref.

6). The management practices described by UM give reasonable assurance that it will continue to be responsible for overall supervision, compliance with regulations, and the health and safety of the public. Therefore, the staff concludes that the proposed project management structure is acceptable.

The prime contractor is an integral part of the organization. The licensee intends to choose the prime contractor using the selection criteria presented above. The staff has reviewed these criteria, which cover all skill areas necessary for successful decommissioning project management and performance. Therefore, the staff concludes there is reasonable assurance that the licensee will select a prime contractor with adequate qualifications to support safe decommissioning of the FNR.

The staff reviewed and compared the licensees organizational and control structures. Based on this review, the staff concludes that the licensee has in place an acceptable organizational structure to safely control the decontamination and dismantlement of the FNR.

3.2.2 Occupational and Public Health and Safety 3.2.2.1 Radiation Protection 3.2.2.1.1 ALARA Program The licensee committed to control decommissioning in accordance with the enhanced requirements of a health physics program and incorporate provisions for reducing individual and collective radiological exposures to ALARA levels. The RSO is responsible for ensuring the establishment of procedures and practices. The FSS plan will fully discuss the implementation of the ALARA requirement of 10 CFR 20.1402 (discussed in Sections 2.1.5 and 4.0 of the DP) to ensure that the ALARA principle is applied to radiation exposures to the public and facility personnel.

3.2.2.1.2 Methods for Occupational Exposure Reduction The licensee presented various methods that will be implemented during the decommissioning project work to ensure that occupational exposure to radioactive materials is minimized. The methods include the use of RWPs, special equipment, techniques, and other practices as described in the DP. RWPs for jobs with low dose commitments will require approval at the health physics technician or health physics supervisory level. The RSO must approve RWPs for jobs with potentially high dose commitment or significant radiological hazards.

The health physics organization will ensure that radiation, surface radioactivity, and airborne surveys are performed as required to define and document the radiological conditions for each job. The licensee committed to instituting process or other engineering controls, as the preferred methods, to maintain exposures to radiation and radioactive materials to ALARA levels. These processes/engineering controls include use of the following:

• shielding to reduce the intensity of external sources of radiation

• containment and confinement structures to prevent/reduce the potential for generating airborne radioactivity

• ventilation systems to remove airborne radioactivity from the work environment In addition to these ALARA measures, Section 3.1.1.1 of the DP discusses other controls that will be employed during decommissioning of the FNR.

3.2.2.1.3 Control and Storage of Radioactive Materials The licensee will continue to rely on the existing health physics program to minimize occupational radiation doses during decommissioning operations in a manner that achieves the following:

• deters the inadvertent release of radioactive materials to uncontrolled areas

• ensures that personnel are not inadvertently exposed to licensed radioactive materials

• minimizes the volume of radioactive waste generated during licensed activities 3.2.2.1.4 Conclusions The licensee has had extensive experience in radiation protection that is directly applicable to decommissioning while operating the reactor facility. The prime contractor will provide further experience and resources under the direction of UM. Based on the review of the DP, the staff concludes that the licensees ALARA program is acceptable.

3.2.2.2 Health Physics Program The DP notes that the existing health physics program at the FNR will remain under the control and authority of UM. Furthermore, the licensee will revise the health physics program as necessary to ensure that it will continue to satisfy the following radiation protection program commitments during decommissioning:

• minimize the radiological impacts to workers, the public, and the environment

• monitor radiation level and radioactive materials

• control distribution and releases of radioactive materials

• maintain potential exposures to the public and occupational radiation exposure to individuals within the limits of 10 CFR Part 20 and at ALARA levels 3.2.2.2.1 Radiation Exposure UM management committed to minimize exposure of individuals to radiation or radioactive materials to ALARA levels. To support this commitment, the licensee will subject individuals conducting decommissioning activities to administrative controls for radiation exposure, which will be based on the requirements contained in 10 CFR Part 20 and may be used to ensure compliance with the annual dose limits and for maintaining exposures at ALARA levels.

The following administrative limits apply to FNR decommissioning activities:

• employees and contractors total effective dose equivalent (TEDE) less than or equal to 2.0 rem/year total organ dose equivalent less than or equal to 2.0 rem/year lens of the eye dose equivalent less than or equal to 2.0 rem/year shallow dose equivalent less than or equal to 2.0 rem/year

• embryo/fetus (declared pregnant worker exposure)

TEDE less than 0.1 rem over the duration of the pregnancy

• visitor, member of the UM community, and member of the public TEDE less than 0.05 rem/year The licensee has written procedures that define administrative limits that are established at levels less that the allowable occupational dose limits specified in 10 CFR 20.1201, Occupational Dose Limits for Adults. Furthermore, prior authorization to exceed these administrative limits for any radiation worker will be obtained, in writing, from the licensees RSO.

The licensee will perform personnel monitoring of occupational radiation exposure from external sources through the use of individual monitoring devices as required by 10 CFR 20.1502, Conditions Requiring Individual Monitoring of External and Internal Occupational Dose. The licensee commits to, at a minimum, on an annual basis, or whenever changes in worker exposures warrant, performing an external exposure evaluation to ensure that personnel monitoring of occupational radiation exposure from external sources is in compliance with 10 CFR 20.1502(a). Dosimeters that require processing (e.g., thermoluminescent or optically stimulated luminescence dosimeters) will be provided by UM and will be processed by a dosimetry processor accredited by the National Voluntary Laboratory Accreditation Program.

The licensee will determine occupational internal exposure from licensed radioactive materials to an individual through monitoring of the quantities of licensed materials in the air collected through air samples, in vitro or in vivo bioassay techniques, or a combination of air monitoring and bioassay as allowed by 10 CFR 20.1204, Determination of Internal Exposure, and required by 10 CFR 20.1502 (b). If respiratory protection equipment is used for protection against airborne radioactive material, then the licensee will evaluate the actual intakes, taking into account the protection factors assigned to the type of respiratory protection employed as allowed by 10 CFR 20.1204. To ensure compliance with 10 CFR 20.1502(b), bioassay for intakes of licensed materials may be performed for licensee personnel with the greatest potential for intake at a sample frequency appropriate for the pulmonary retention class (days, weeks, years).

When exiting restricted areas that have known removable contamination or the potential for removable contamination, site personnel will monitor their hands and feet for contamination in accordance with internal procedures. If contamination is detected, then the site personnel will check the exposed areas of the body and clothing. Site personnel leaving potentially contaminated areas will periodically monitor their hands and feet for contamination, consistent with the nature and quantity of the radioactive materials present.

The licensee will continue to measure the concentrations of radioactive material released from the facility in gaseous effluents. The dilution factor of 400, taken from previous safety analyses submitted to the NRC and contained in the TSs, continues to apply to the FNR exhaust and the PML stack exhausts. UM may also use other options for showing compliance with the annual dose limit to an individual member of the public from concentrations of radioactive material released from the facility in gaseous effluents, as allowed by 10 CFR 20.1302, Compliance with Dose Limits for Individual Members of the Public.

To ensure compliance with the requirements of 10 CFR Part 20, the licensee will continue to measure the concentrations of radioactive material released from its facility in liquid effluents.

UM may also use other options for showing compliance with the annual dose limit to an individual member of the public from concentrations of radioactive material released from the facility in liquid effluents as allowed by 10 CFR 20.1302.

3.2.2.2.2 Surveys and Monitoring The licensee will perform radiation surveys and monitoring in accordance with the existing radiation protection program and as necessary to support work activities in areas with the potential for exposure to radiation or radioactive materials. The licensee will assess the effectiveness of controls to minimize or eliminate radiation exposures in the following two ways:

(1) direct measurement of the external radiation or the radioactive material intake an individual receives (2) measurement of the radiological conditions in the area(s) occupied by the individual Levels and extent of direct radiation and radioactive materials in any work area will be measured and assessed in accordance with the licensees health physics program. These measurements will include, as a minimum, the following:

I. direct dose rate measurements II. surface contamination measurements (fixed and removable)

III. airborne radioactive material measurements The licensee will ensure that instruments and equipment used for these measurements are calibrated for the radiation type to be measured on frequencies as listed in Section 3.1.2.4 of the DP.

3.2.2.2.3 Exposure Control The licensee defines restricted areas based on the known or suspected hazard potential from radiation sources that have been defined from measurement or inferred from process knowledge. Radiation exposures to an individual entering such an area may be assessed from any combination of the following:

IV. direct radiation V. surface contamination (fixed and removable)

VI. airborne contamination 3.2.2.2.4 Control of Exposure to Direct Radiation Control of exposure to individuals from direct radiation is based on two elements, as defined in the DP:

(1) measurement and assessment of the location and strength of the radiation sources (2) control of the individuals access to those radiation sources Routine monitoring of the levels and extent of radiation and radioactive materials is a key part of the licensees health physics program. The program also involves measuring and assessing the levels and extent of direct radiation and radioactive materials in work areas. These measurements include direct dose rate, surface contamination, and airborne radioactive material measurements.

Before defining control requirements for limiting direct radiation exposure to individuals, the licensee will determine the location of the radiation sources and the magnitude of the radiation.

Direct radiation exposure measurements will be made at the time of decommissioning, concentrating on areas identified as having a worker exposure potential. This survey work also will include specific areas or systems identified by the licensee during work planning before project startup.

Based on this measurement and data assessment, the licensee will establish shielding, or barriers that restrict access to sources of radiation. The licensee will install postings at access points through those barriers based on the potential exposures that an individual could receive upon entry through the access points or along external surfaces of the barrier, in accordance with regulatory requirements.

3.2.2.2.5 Control of Exposure to Surface Contamination The licensee will control exposure to individuals from surfaces contaminated with radioactive material either by prior decontamination or by using protective equipment for personnel to minimize or limit exposure to the surface material.

Prior decontamination for planned work activities is the licensees preferred method of contamination control. However, the licensee will evaluate this practice for ALARA considerations to ensure that exposures resulting from the decontamination/removal do not offset exposure savings for the planned work activities.

The licensee may need to establish controlled surface contamination areas because the contaminants present at the FNR are primarily beta-gamma emitting activation and fission products. The licensee will use administrative control postings for contamination areas and high contamination areas as follows:

VII. contamination areaan area where surface contamination levels exceed the requirements for unrestricted release of a surface, but are less than 100 times the surface values in Table 3-1 of the DP VIII. high contamination areaan area where surface contamination levels exceed 100 times the surface values in Table 3-1 of the DP When decontamination is impractical or ineffective, personal protective equipment (PPE) will be used to protect individuals from potential radiation exposures attributable to surface contamination. The licensee will consider radiological conditions, type of work to be performed, potentially stressful environmental conditions, physical condition of surfaces, and duration of the activity in determining the appropriate PPE.

If the potential for exposure to airborne radioactivity at levels in excess of 12 derived air concentration-hours in a workweek is encountered, the licensee will require workers to don full-face respirators for work activities that will be performed in these areas.

As appropriate, the licensee will employ contamination control measures that include, but are not limited to, the following:

IX. local containment barriers such as designed barriers, glove bags, containers, and plastic bags to prevent the spread of radioactive material X. physical barriers such as Herculite sheeting, strippable paint, tacky-mat step-off pads, absorbent pads, and drip funnels to limit contamination spread 3.2.2.2.6 Control of Exposure to Airborne Contamination If air monitoring results indicate levels of airborne radioactive materials in excess of NRC-prescribed levels, the licensee will post the area as an airborne radioactivity area at access points, per the definition in 10 CFR 20.1003, Definitions.

When it is not practical to employ the engineering controls described previously, or when these controls are not sufficient to maintain the airborne radioactivity levels below those defining an airborne radioactivity area, the licensee will then require the use of respiratory protection equipment for individuals entering this work environment. When respiratory protection is required, it will be as described in a respiratory protection program satisfying the requirements of 10 CFR Part 20, Subpart H, Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas. The licensees program will include worker training and medical qualification requirements for use and descriptions of the following:

XI. respiratory protection equipment to be used XII. air monitoring requirements to support the use XIII. bioassay program to evaluate the effectiveness of use XIV. equipment cleaning, testing, and maintenance requirements 3.2.2.2.7 Radiation Monitoring Equipment The licensee will maintain a sufficient inventory and variety of instrumentation onsite to facilitate effective measurement of radiological conditions and control of worker exposure consistent with ALARA principles and to evaluate the suitability of materials for release for unrestricted use.

The licensee will employ radiation monitoring equipment capable of measuring the range of dose rates and radioactivity concentrations expected to be encountered during remediation and decontamination activities to the minimum values required for release of materials for unrestricted release.

The licensee committed to calibrate radiation monitoring equipment at the intervals prescribed by the manufacturerannually, or prior to use as discussed in Table 3-2 of the DP. Radiation monitoring equipment will be calibrated using standards traceable to the National Institute of Standards and Technology (NIST). Calibration information will be clearly marked on the instrument. In addition, survey instruments and equipment will be operationally tested daily when in use.

3.2.2.2.8 Conclusions The licensee has a mature health physics program capable of protecting workers and minimizing the levels of radiation exposures that may be encountered during decommissioning of the FNR. As such, the NRC staff finds that there is reasonable assurance that the implementation of the procedures and guidance of the health physics and ALARA programs will minimize the radiation exposure of workers and the public. The staff concludes that the licensees health physics program is acceptable and meets the requirements in 10 CFR 20.1101, Radiation Protection Programs.

Based on the review of the respiratory protection program proposed in the DP, the staff concludes that the licensee has the necessary organizational structure and management controls to establish and maintain a program that meets the requirements of 10 CFR Part 20, Subpart H.

3.2.2.3 Control of Radioactive Materials The licensee will survey all materials leaving a restricted area to ensure that such equipment, materials, and items do not contain detectable quantities of radioactivity. The licensees surveys will incorporate the guidance in NRC Circular No. 81-07, Control of Radioactively Contaminated Material, dated May 14, 1961 (Ref. 7), and Information Notice No. 85-92, Surveys of Wastes Before Disposal from Nuclear Reactor Facilities, dated December 2, 1985 (Ref. 8).

For items that may be contaminated with beta-gamma emitting activation and fission products, the licensee will use the following survey methods:

XV. materials and equipmentdirect frisking with a portable Geiger-Mueller detector (e.g.,

Ludlum Model 44-9, Eberline Model HP-210, or equivalent) having a minimum level of detection above background of less than or equal to 5000 disintegrations per minute (dpm) per 100 square centimeter (cm2)

XVI. smear samplesanalysis with a Geiger-Mueller detector (e.g., Ludlum Model 44-9, Eberline Model HP-210, or equivalent) having a minimum detection level above background of less than or equal to 1000 dpm per 100 cm2 XVII. bulk materials (e.g., sand and soil)analysis of representative sample(s) using a high-resolution gamma spectroscopy system having a lower limit of detection above background of less than or equal to 0.18 picocurie (pCi) per gram for cesium (Cs)-137 XVIII. background-equivalent gamma activityan unshielded gamma ray dose measured 1 meter from any surface, not to exceed 5 microrem per hour above background The licensee may develop additional methods for release of surface contaminated materials, which would be subject to the minimum detection levels in Table 3-3 of the DP. Detection sensitivities of instruments and techniques may be determined using the guidance contained in the NUREG-1575, Revision 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), issued August 2000 (Ref. 9), and NUREG/CR-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, issued June 1997 (Ref. 10). The licensee may relocate equipment and materials to areas of lower ambient background for the conduct of release surveys. The licensee will release materials only if a survey using a method identified above does not identify any discernable radioactivity above background from licensed materials.

In evaluating equipment and materials for fixed or smearable licensed radioactive materials, the licensee will not release items painted with other than the original manufacturers paint unless clear process knowledge demonstrates that the paint was applied to a clean surface containing no discernable radioactivity from licensed materials prior to its use in a restricted area.

Following the satisfactory completion of a survey satisfying the requirements listed above, the licensees RSO may approve the release of items for which it cannot be demonstrated that paint was applied to a clean surface.

If the potential exists for contamination on inaccessible surfaces, the licensee will assume the equipment to be internally contaminated unless (1) the equipment is dismantled allowing access for surveys, (2) appropriate tool or pipe monitors satisfying the survey requirements listed above are used to provide confidence that no licensed radioactive materials are present, or (3) it may readily be concluded that surveys from accessible areas are representative of the inaccessible surfaces (i.e., surveying the internal surface of both ends of a pipe from a nonradioactive process system with cotton swabs would be representative of the inaccessible areas).

Personal effects (e.g., notebooks, pens, flashlights) that are hand-carried into a restricted area are subject to the same survey requirements as the individual possessing the item.

The licensee may transfer licensed radioactive materials to other locations within the control of UM as allowed by appropriate radioactive material licenses issued by the NRC. The licensee may transfer licensed radioactive materials to other locations outside UM that possess the appropriate radioactive material licenses issued by the NRC, an Agreement State, or are otherwise authorized to possess such radioactive material (e.g., DOE sites, foreign research reactors).

The licensee will revise its existing health physics program (described in Section 3.2.2.2) revised as needed in accordance with the internal approval and change control provisions discussed in Section 2.4 of the DP.

3.2.2.3.1 Conclusions The NRC staff finds that the licensee has an adequate program for assessing whether equipment, tools, materials, and items contain detectable quantities of radioactivity prior to their unrestricted release. The NRC staff bases its conclusion on the licensees commitment to adhere to NRC Circular No. 81-07 for making this assessment.

Furthermore, based on commitments from the licensee, volumetric releases of bulk materials for unrestricted release will be limited solely to soil and sediment, but will not include rubblized concrete or other similar manmade items. Should the licensee elect to seek approval to release such manmade items in the future, the NRC staff noted that an exemption under 10 CFR 20.2002, Method for Obtaining Approval of Proposed Disposal Procedures, would be required based on a site-specific radiological dose assessment reflective of the site where such items would ultimately be dispositioned.

Based on information and commitments provided by the licensee, the NRC staff concludes that the licensee has an adequate health physics program for ensuring that equipment, tools, materials, and items do not contain detectable quantities of radioactivity prior to their unrestricted release in support of the decommissioning of the FNR.

3.2.2.4 Dose Estimates The licensee estimated the total occupational exposure to complete the FNR decommissioning project to be 4.8 person-rem. The licensee based this dose estimate on characterization data and professional judgment that took into account the individual work activity durations and work crew sizes estimated by the UM contractor.

While the dose estimate provided in the DP is for planning purposes only, the licensee will develop detailed exposure estimates and exposure controls in accordance with the requirements of the ALARA program. The licensee will perform the actual estimate of exposure that may be incurred in the course of decommissioning during detailed planning of the decommissioning activities.

The licensee estimated that the dose estimate to members of the public as a result of decommissioning activities will be negligible. The licensee based this conclusion on the fact that the area immediately surrounding the facility is under UM control and because the area where decommissioning activities are taking place is fully contained within the facility (with the exception of loading and unloading shipments of equipment and radioactive materials). This is consistent with the negligible (less that 0.1 person-rem) dose estimate provided for the reference research reactor in NUREG-0586, Final Generic Environmental Impact Statement on the Decommissioning of Nuclear Facilities, issued 1988 (Ref. 11).

3.2.2.4.1 Conclusions Based on the NRC staffs review of information provided by the licensee, the licensees estimates for occupational and public dose during decommissioning activities are reasonable and well below the radiation dose levels permissible to workers and members of the public as specified in 10 CFR Part 20. Therefore, the NRC staff concludes that the FNR can be safely decommissioned to levels allowed under such Federal regulations.

3.2.2.5 Radioactive Waste Management Section 3.5.3 of this document discusses quality assurance provisions for the transport of licensed radioactive material.

3.2.2.5.1 Fuel Removal To support issuance of a possession-only license, the NRC approved a revised TS that prohibits the licensee from operating the FNR and possessing reactor fuel. The agency based this licensing action upon the licensees decision to remove all irradiated and unirradiated reactor fuel from the site. All irradiated reactor fuel was removed from the FNR and returned to the DOEs Savannah River Site between October and December 2003. In addition, the licensee returned all unirradiated reactor fuel to DOE through BWXT Technologies in August 2003.

3.2.2.5.2 Radioactive Waste Processing The licensee stated that in general, system components will not be decontaminated onsite.

Slightly contaminated items may be decontaminated onsite if it is determined that a component or portion of a component can be safely and economically decontaminated. Onsite personnel, including staff, contractors, or specialty contractors will decontaminate such items using techniques and materials within the capabilities of those personnel as determined by UM.

Experienced offsite vendor(s) may also be used to decontaminate the components if they can be safely and economically decontaminated as determined by UM. Currently, the licensee intends dismantle the contaminated piping systems and dispose of the material as radioactive waste or to decontaminate and free release those materials.

The licensees decommissioning of the FNR will result in the generation of solid and liquid low-level radioactive waste, mixed waste, and hazardous waste. Solid radioactive wastes include neutron-activated materials, contaminated materials remaining in the reactor building, and those items necessarily contaminated onsite during the remediation activities. The licensee anticipates limited, if any, soil remediation that would result in generating solid radioactive wastes. Liquid low-level radioactive waste includes the water in the reactor pool and the associated piping as well as contaminated water generated during remediation activities. The licensee stated that no gaseous radioactive waste exists because the reactor has been shut down for more than 9 months and all radioactive gases have since decayed.

The licensee will perform handling, staging, and shipping of packaged radioactive waste in accordance with all applicable Federal, State, and local regulatory requirements.

The majority of the solid waste expected to be generated during decommissioning of the FNR is expected to consist of Class A low-level radioactive waste. Onsite radioactive waste processing will include waste minimization, volume reduction, segregation, characterization, neutralization, stabilization, solidification, and packaging. Wastes may be shipped to a licensed processing facility for survey and release or decontamination and release, or may be disposed of directly at a licensed facility in accordance with its radioactive material license. A manifest consistent with the proper waste classification will accompany each shipment of radioactive waste, as specified in Section I, Manifest, of Appendix G, Requirements for Transfers of Low-Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests, to 10 CFR Part 20.

3.2.2.5.3 Radioactive Waste Disposal Low-Level Liquid Radioactive Waste The licensee plans to dispose of approximately 50,000 gallons of low-level radioactively contaminated water currently contained in the reactor pool and associated piping by treatment and discharge to the public sewer system. The City of Ann Arbor operates the sewer system in accordance with Federal, State, and local regulatory requirements. Any additional radioactively contaminated water generated during remediation activities may also be similarly discharged in accordance with all applicable regulatory requirements.

The licensee will monitor and process the low-level radioactively contaminated water from the reactor pool and associated piping using techniques consistent with the licensees health physics and ALARA programs.

The liquid waste generated during licensee remediation activities will be monitored and processed prior to discharge. During demolition activities, installed plant equipment used to process liquid radioactive waste may be removed. Therefore, temporary filtration units or demineralizers may be used as the primary means of treatment. Any temporary liquid treatment system necessary to ensure that disposal requirements are met will be connected to tanks for storage of processed water prior to discharge. Once the licensee verifies that the stored processed water meets the allowable discharge limits specified in the TSs, the water may be subsequently released. The existing effluent monitoring instrumentation will be used to monitor discharges of liquid effluent as required and to demonstrate compliance with the TSs and other applicable regulations.

Makeup water used for flushing will typically originate from the existing potable water supply.

The effluent stream(s) from such activities will be processed as above, by filtration and demineralization.

The licensee will implement measures to treat liquid radioactive waste to maintain worker radiation doses within the required regulatory limits during decommissioning of the FNR. As such, filters will be replaced as appropriate and system components will be positioned or shielded.

If radioactively contaminated water is discharged to the sanitary sewer, the discharge piping will be resurveyed and remediated as necessary.

The DP details the following disposal options:

XIX. Low-level radioactively contaminated water may be evaporated onsite. For such activities, the facilitys effluent monitoring system is equipped to monitor airborne radioactive effluents in accordance with the TSs and applicable regulations.

XX. The licensee may choose to use a licensed radioactive waste processor(s) to provide specialized services for reducing the volume of or treating radioactive liquid waste.

Such services may include demineralization, direct incineration, ground application, evaporation, and survey and release. UM may also elect to transfer all or some of the liquid radioactive waste from decommissioning to a licensed waste processor.

XXI. Currently, the licensee does not plan to use chelating agents in any chemical decontamination activities for FNR systems or structures. Radioactive wastes containing chelating agents will be generated only if necessary and will be minimized to the fullest extent possible.

XXII. The licensee will return tritium-loaded heavy water, owned by DOE, to the DOE Savannah River Site for processing and reuse.

Solid Radioactive Waste While the majority of solid waste generated during the decontamination and dismantlement of activated and contaminated systems, structures, or components is expected to consist of Class A low-level radioactive waste, the licensee noted that information on the estimated curie (Ci) content and waste volume for this decommissioning project is extremely limited at this time.

Additional information is therefore required to determine the specific waste classification. The estimates of waste volumes are conservative and do not account for any volume reduction techniques. In addition, the estimates assume only direct burial rather than allowing for decontamination and possible free release.

The licensee is planning a number of measures to reduce the volume of solid radioactive waste that will require disposal at a licensed burial facility. The primary components of the solid waste to be generated by the decommissioning of the FNR facility are expected to be disposed as summarized below and in Section 3.2.4 of the DP:

I. Irradiated reactor hardware may require size reduction to facilitate loading. Irradiated reactor hardware will be loaded into a high-integrity container (HIC) or liner, then placed in an approved, shielded shipping cask for transport and subsequent direct burial at the licensed land disposal facility in Barnwell, South Carolina. The current estimate for the volume of irradiated reactor hardware requiring burial at Barnwell is 300 cubic feet.

II. The contaminated systems piping and equipment will be segmented. As cuts are made, a suitable cover will be placed on open ends to preclude the spread of contamination.

Material that can be economically dismantled and decontaminated will be appropriately handled onsite or sent to a vendor facility for decontamination. Material that cannot be economically decontaminated will be placed in proper disposal containers (e.g., low specific activity (LSA) containers) and sent to an appropriate processor or burial facility. The licensee expects approximately 5300 cubic feet of activated or contaminated material to be generated for processing or disposal.

III. Activated or contaminated concrete removed in large sections will be packaged as LSA material in approved shipping containers for direct shipment to the licensed land disposal facility operated by Envirocare of Utah, Inc. An estimated 5200 cubic feet of activated or contaminated concrete, two-thirds of the concrete comprising the reactor pool, will require disposal in this manner.

IV. Dry Active Waste (DAW) consisting of contaminated paper, plastic, coveralls, and similar items will be packaged as LSA material in approved shipping containers. The licensee will ship uncompacted DAW to an offsite vendor for volume reduction and processing if supported by ALARA and cost considerations. When feasible, DAW will be used to fill void space in other radioactive waste shipping containers; otherwise, it may be shipped for direct burial. An estimated 300 cubic feet of DAW will require transfer to a licensed waste disposal facility for postprocessing and disposal.

V. Engineering controls such as high-efficiency particulate air (HEPA)-filtered ventilation will be required to capture potential airborne contaminants. Spent HEPA filters will be changed out and treated as DAW. An estimated 25 cubic feet of contaminated filter media will require transfer to a licensed waste disposal facility for postprocessing and disposal.

VI. Radioactive waste treatment systems will be required to process the liquid waste stream resulting from various decommissioning activities as described above. The licensee will use filtration and ion exchange processing to remove residual radioactivity in the water. A vendor or the FNR may supply temporary demineralization and filtration systems. The licensee estimates the volume of spent resins and filters required to process the water to be less than 400 cubic feet. These resins will be transported to a licensed facility for disposal.

VII. The licensee does not expect radioactively contaminated asbestos waste to be present, but it may be identified by decommissioning or preparatory activities.

Asbestos material should be transferred to an offsite, licensed radioactive waste processor for compaction or for survey and release. Large items containing asbestos waste may require size reduction before transfer to the offsite, licensed radioactive waste processor.

VIII. The only known mixed waste at the FNR is from lead shielding, possibly lead paint, and cadmium. The FNR has approximately 13,000 pounds of contaminated lead, 1,600 pounds of activated lead, 400 pounds of contaminated cadmium, and 20 pounds of activated cadmium. A vendor will encapsulate or otherwise treat these materials for ultimate disposal or recycle. The objective of UM is to generate no new mixed waste during decommissioning activities. Procedures currently in place for hazardous and radiological waste management are sufficient to provide the assurance that waste will not be generated arbitrarily and that generated wastes will be disposed of properly.

3.2.2.5.4 Method of Estimating Types, Amounts, and Radionuclide Concentrations of Radioactive Waste Generated during Decommissioning The licensee will derive an estimate of total radioactivity present in systems, structures, or components directly from field radiological measurements, supplemented by analytical data or through computational estimates, as follows:

IX. sampling volumetric material to establish ratios of radionuclides present in a structure or component X. direct measurement using sodium iodide (NaI), high-purity germanium, or other detectors to analyze the gamma spectrum being emitted to identify specific isotopes, establish ratios of isotopes, or to fully quantify isotopes XI. direct measurement of dose rates to support computational methodologies for the determination of radionuclides present XII. direct measurement of similar items for extrapolation via computational methods for inaccessible components or structures Estimates of the radionuclide concentration in irradiated items may be based on the constituent elements of the material in question and by calculating the duration of exposure and the energies of the incident neutrons. The licensee will use radiological surveys to determine the activity present within internally contaminated piping and on structures.

3.2.2.5.5 Conclusions Based on the review of the licensees program as described in the DP and the licensees experience, the NRC staff concludes that the licensees proposed radioactive waste management plans for the UM decommissioning project are acceptable and will conform to NRC regulations.

3.2.3 Training Program Because decommissioning activities are much different from typical FNR operations, the licensee committed to conduct special training for the existing FNR operations staff and the decommissioning personnel. Individuals (employees, contractors, and visitors) who require access to the work areas or radiologically restricted area will receive training commensurate with the applicable regulatory requirements (i.e., 10 CFR Part 19, Notices, Instructions and Reports to Workers: Inspections and Investigations) for the potential hazards to which they may be exposed. Individuals will also receive continued training, as necessary, to ensure that job proficiency is maintained.

Personnel will be qualified for their assigned duties prior to performing such work or will be under the direct supervision of a qualified employee. Personnel performing special processes will be qualified according to specific codes and standards and/or in accordance with national consensus documents. Qualification will include proficiency demonstrated by each individual prior to performing work and periodically assessed throughout the duration of the project.

Qualification also will be demonstrated when required by the designated codes or standards.

The licensee will maintain training records that include the trainees name, dates of training, types of training, test results, protective equipment use authorizations, and instructors name.

Care will be taken to ensure that properly qualified instructors conduct all training. As the primary criterion, persons responsible for presenting training should have knowledge and experience in the process or subject matter. It is desirable that trainers also have the presentation skills or classroom conduct appropriate to the level of the training being presented.

For those with limited experience in conducting training, early instruction should be monitored and feedback should be provided.

The licensee provided examples of the various types of training programs applicable to decommissioning activities::

I. general employee traininggeneral training for emergency response, spill response, alarms, alarm response, communication systems and channels, waste management, and waste minimization II. radiation safety training:

general radiological trainingtraining for personnel who are required to enter radiological restricted areas, with the exception of visitors and infrequent support personnel, but are not authorized to perform hands-on radiological work radiological worker trainingtraining for personnel who require unescorted access to radiological restricted areas and who are authorized to perform radiological job functions core trainingmay be accomplished under any program that meets basic requirements site-specific traininggiven to all personnel refresher traininggiven annually to all personnel

• hazardous waste operations and emergency responsetraining for personnel engaged in hazardous substance removal or other activities that potentially expose them to hazardous substances and health hazards, which satisfies 29 CFR 1910.120, Hazardous Waste Operations and Emergency Response

• respirator training and fit testingtraining, medical qualification, and fit testing for each person who wears a tightly fitting respirator that satisfies the requirements of 10 CFR Part 20, Subpart H, and Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection (Ref. 12)

• Department of Transportation hazardous materials employee trainingtraining as required by 49 CFR Part 172, Hazardous Materials Table, Special Provisions, Hazardous Materials Communications, Emergency Response Information, and Training Requirements, Subpart H, Training, provided to all personnel involved in the loading, unloading, or handling of hazardous materials, preparing hazardous materials for transportation (including packaging and preparation of manifests), or responsible for the transportation of radioactive materials or operation of a vehicle used to transport hazardous materials (49 CFR 171.8, Definitions and Abbreviations)

• security requirements for offerors and transporters of hazardous materialstraining for in the facilitys security plan that satisfies the requirements of 49 CFR Part 172 for all personnel involved in the offering of placarded quantities shipments of hazardous materials

• hazard communication trainingtraining covering, at minimum, the proper use of materials, the required PPE, and the emergency procedures associated with these materials for all personnel on the hazardous chemicals in their work area, as required by 29 CFR 1910.1200(h), including update training whenever a new physical or health hazard is introduced into their work area

• hearing conservation trainingtraining on the effects of noise on hearing and the purposes, advantages, disadvantages, and attenuation of various types of hearing protective devices

• permit-required confined space entry trainingtraining for personnel if entry into confined spaces is to be performed

• lockout/tagout trainingtraining for hazardous energy control

• trenching and excavation trainingtraining for the purpose of determining the safety and stability of excavations

• fire watch trainingtraining on the proper selection, use, and application of extinguishing agents; characteristics and classification of fires

• asbestos abatement trainingtraining on requirements, potential health effects, and controls for asbestos abatement

• torch/plasma arc cutting, welding, and open flame trainingstraining in the use of, and understanding the reasons for, protective clothing and equipment, including the need for flame-resistant clothing

• tailgate trainingroutine, short training, given usually at the beginning or end of a regular workforce briefing, intended to provide a brief review of a safety or programmatic topic, which is applicable to current work activities

• other specific mandated trainingany other training that may be required by the standards specific to the Michigan Occupational Safety and Health Act of 1974 (MIOSHA) or applicable standards before initiating work that may fall within the scope of decommissioning 3.2.3.1 Conclusions Based on the review of the licensees training program as described in the DP, the staff concludes that the licensees training program is acceptable. The licensee also recognized that specific training would be required to reflect the unique hazards associated with decommissioning operations. While the NRC does not regulate nonradiological hazards as specified in the Atomic Energy Act, the licensee is aware that personnel involved with decommissioning activities would be subject to training requirements administered by other Federal, State, and local government agencies.

3.2.4 General Industrial Safety Program The licensee stated that the RSO, with the cooperation of the full project management team, will be responsible for ensuring that the occupational health and safety requirements for project personnel are met, primarily in terms of compliance with the Occupational Safety and Health Act of 1973 and MIOSHA. Specific responsibilities include establishing training requirements for general safe work practices, reviewing plans and procedures to verify adequate coverage of industrial hygiene and safety requirements, conducting periodic inspections of work areas and activities to identify and correct any unsafe conditions and work practices, coordinating industrial hygiene services as required, and advising the Director on industrial hygiene and safety matters and on the results of periodic safety inspections.

All personnel working on the FNR decommissioning project will receive health and safety training in order to recognize and understand the potential risks to personnel health and safety associated with the work at the FNR. The health and safety training also ensures compliance with the applicable regulatory requirements. Personnel will be trained on the plans, procedures, and operation of equipment to conduct work safely on the FNR decommissioning project.

The implementation of occupational health and safety requirements for activities involving potential hazards that may be encountered during decommissioning will be evaluated through the use of a JHA. Each JHA will identify all hazards associated with the activity (e.g., fall protection, hot work, confined space). The licensee will prepare a procedure implementing the JHA that will be subject to the approval requirements discussed in Section 2.4 of the DP. The JHA allows the project management, project staff, contractor staff, and UM industrial safety personnel (through the RSO or reactor manager) to specify the controls and processes necessary to protect the safety of individual workers, the UM community, and the public. The JHA will act in concert with the RWP, if required, to complete the protection program. A representative of the UM industrial safety staff, the RSO, or the reactor manager will approve the JHAs. In their absence, the RSO and the reactor manager can delegate this approval authority.

3.2.4.1 Conclusions Based on the review of the licensees proposed industrial safety program as outlined in the DP, the staff concludes the program is acceptable.

3.2.5 Radiological Accident Analyses The licensee evaluated radiological accidents that could potentially occur during decommissioning of the FNR. This accident analysis considered areas that contain the highest inventories of radioactive material expected to be present during the decommissioning of the FNR. The results of this analysis adequately bounded the radiological impacts that could reasonably occur during decommissioning. As such, a fire, a pool leak, and a tritium-loaded heavy-water spill were the radiological accidents considered to present the highest potential consequences.

3.2.5.1 Fire The licensee considered the consequences of a fire during decommissioning of the FNR and did not find them to be significantly different than the consequences of a fire during reactor operations. The majority of the materials of construction present in the FNR are metals, concrete, or similar noncombustible materials. Upon termination of reactor operation, most of the combustible materials required for reactor operations were removed from the reactor building to further reduce the potential consequences of a fire. The licensee concludes that it is highly unlikely that a fire would start or that a fire could become intense enough to ignite these types of materials (including other combustible materials such as rags, wipes, and anticontamination clothing), and thus result in the release of radioactive material.

The licensee stated that dry radioactive waste is normally collected in metal pails with lids located throughout the facility. Once full, the dry waste is normally transferred into 55-gallon drums meeting the strong-tight requirement for shipment to a licensed waste processor. Small quantities of dry radioactive waste requiring special handling or segregation are stored in plastic 5-gallon pails. The licensee stated that this practice limits the volume of dry radioactive waste that could be ignited in a fire event to a few pounds and serves to lower the potential for a fire to consume additional waste collections. The licensee contends that any fire involving dry radioactive waste would be limited to a few microcuries of radioactivity from radionuclides contained in the list of expected radionuclides (refer to Table 2-4 of the DP).

During a fire involving dry radioactive waste, the emission of airborne radioactivity from the FNR exhaust stack would continue unless operator action is taken, or upon automatic closure of the ventilation dampers when the radioactivity levels exceed 1 millirem (mrem) per hour at the building exhaust radiation monitor (required by the TSs). The licensee stated that for the purposes of the evaluation, the ventilation dampers were assumed to remain open, and an exhaust stack dilution factor of 400 and an emission rate of a minimum of 8000 cubic feet per minute up the FNR exhaust stack was assumed, for a duration of 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />.

Table 3-1 presents the emissions of individual radionuclides that could be released to the environment resulting from a fire without exceeding the airborne effluent concentration (AEC) limits for a full year as specified in Table 2 of Appendix B, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage, to 10 CFR Part 20.

Table 3-1 Quantities of Individual Expected Radionuclides Producing the Emission of the AEC during an 8-Hour Fire Nuclide Individual Quantity1 Antimony-125 W class 130 mCi Bismuth-210m D class 0.4 mCi Cadmium-109 W class 8.7 mCi Carbon-14 (Monoxide) 86 Ci Cesium-134 D class 8.7 mCi Cesium-137 D class 8.7 mCi Cobalt-60 W class 8.6 Ci Europium-152 W, All classes 390 mCi Europium-154 W, All classes 1.3 mCi Iron-55 W class 260 mCi Manganese-54 All classes 44 mCi Nickel-59 W class 430 mCi Nickel-63 W class 173 mCi Scandium-46 Y, All classes 13 mCi Silver 108m W class 17 mCi Silver 110m W class 13 mCi Tritium 4.3 Ci Zinc-65 Y, All compounds 17 mCi 1

Activity = AEC x 28,317 cc/ft3 x 8,000 cfm x 60 min/h x 8 h x 400 Because these quantities of radionuclides are in the mCi range, the licensee stated that they are significantly greater than the levels expected in any localized, individual containers of dry radioactive waste (i.e., rags, wipes, and anticontamination clothing). During a fire involving these types of wastes in which the ventilation system is secured shortly after the initiation of the fire, the licensee described that exposure would be limited to those individuals who provide initial emergency fire suppression activities. As such, the bounding consequence of the most credible fire event involving an individual container would result in a maximum exposure of 50 mrem to a member of the public.

The construction of the reactor building provides three locations where hot gases from a fire would collect. One of these locations is the area above the reactor pool; the other two locations are just below the third floor on both the east and west sides of the reactor pool. The licensee concludes that this inherent design feature aids in the reduction of the concentration of radioactive materials in the breathing space near the first, second, and third floors of the reactor building, and limits the inhalation of radioactive materials of individuals who provide initial fire suppression activities, those individuals who evacuated the facility, and those individuals who are required to reenter the reactor building.

In the event of a fire, individuals present in the facility may make a reasonable attempt to extinguish the fire using the portable extinguishers provided throughout the facility. If the fire cannot be extinguished, the Ann Arbor Fire Department is summoned, as discussed in the FNR Emergency Plan. According to the licensee, it expects only minimal radiological exposures to be incurred by individuals during the short period while attempting to extinguish the fire in the dry radioactive waste or evacuating the area. In addition, fire-fighting personnel responding to a fire potentially involving radioactive materials are trained to use PPE (including adequate respiratory protection, which the licensee concludes would ensure that any internal exposure would be significantly less that the 50 mrem bounding dose analyzed for a member of the general public).

3.2.5.2 Pool Leak In the event of a major leak from the reactor pool, all water loss would be collected by the floor drains or pass through openings in the first floor to the basement of the reactor building. The dimensions of the reactor basement are large enough to allow for the collection of all 50,000 gallons of water from the reactor pool. Loose radioactive contamination in the waters pathway to the reactor basement would be entrained, which the licensee contends should not cause an increase in the content of radioactive materials in the pool water above the levels experienced while the reactor was operating. The licensee concludes that the resulting levels of radioactive material from the evaporation of the water spread over the basement and first floors would be limited to the tritium contained in the water and would be less than the evaporation rates experienced from the surface or the reactor pool while the reactor was operating. The other radionuclides would remain in the facility. The licensee stated that during normal operation of the reactor pool, prior to reactor shutdown, the 240 square feet of the pool's surface was maintained between 90 EF and 116 EF with an estimated evaporative loss rate of 4 gallons per hour.

3.2.5.3 Tritium-Loaded Heavy-Water Spill The consequences of a spill from a 55-gallon drum of tritium-loaded heavy water would be the emission of tritium via the FNR exhaust stack. Taking credit for the FNR exhaust stack dilution factor of 400, and assuming the emission of 8000 cubic feet per minute up the FNR exhaust stack, the licensee calculated the emission of tritium from the facility to be 9.1 mCi per hour based upon the AEC limit for a full year as specified in Table 2 of Appendix B to 10 CFR Part 20. The heavy-water reflector contains the most concentrated tritium-loaded heavy water. At 217 Ci of tritium (as of April 2004) in an estimated 50 gallons, the licensee calculated the highest estimated concentration of tritium to be 1.1 mCi/milliliter (mL). Given this concentration, a spill from this tank would require the evaporation rate to be limited to approximately 9 mL per hour if the emission were averaged over an entire year. Any spill of tritium-loaded heavy water could be easily flushed to the floor drains for collection in the hot and cold sumps and eventual collection in the retention tanks. Conservatively, 1 week or less would be needed to clean up the spill or to stop the tritium evaporation. The licensee calculated that the emission rate could increase to 473 mCi/hour over 1 week. This equates to an evaporation rate of 0.473 liters (L) of the tritium-loaded heavy water per hour for the entire week of cleanup activities. The licensee concludes that the emission of tritium at a rate of 473 mCi per hour for the 1 week of cleanup would result in a maximum exposure to an individual of 50 mrem.

In the event of a spill of the heavy-water reflector while still in the reactor pool, the licensee stated that the dilution by the water in the pool would decrease the concentration of the tritium in the water source and result in a lower emission rate of tritium from the facility (see License Amendment Nos. 36 and 46).

3.2.5.4 Conclusions The licensee analyzed bounding accidents that may occur during the decommissioning project.

Based on the NRC staffs review of the information provided by the licensee, the radiological consequences for the types of accidents that may potentially occur during decommissioning of the FNR are bounding and within the limits specified in 10 CFR Part 20.

3.3 Decommissioning Activities 3.3.1 Radiological Status of the Facility 3.3.1.1 General The licensee listed potential causes of radioactive contamination in the reactor building from normal operations and routine activities as well as nonroutine occurrences, operations, accidents, and spills. Based on these historical reviews, in addition to characterization surveys of the facility during nearly 50 years of operation of the FNR, the licensee determined that events occurred that led to the radiological contamination of the facility. However, the prompt response and cleanup activities initiated by the facility staff limited contamination in areas not expected to be contaminated by routine operations. Additionally, the licensees practice of periodic monitoring and maintaining contamination action levels between 3 and 10 times background resulted in a limited number of areas where contamination levels are reported to exist above the anticipated release criteria.

3.3.1.2 Principal Radioactive Components The information obtained by the licensee indicates that the radioactive portions of the facility are primarily confined to the reactor internals and reactor pool. The licensee estimated the radioactivity inventory by considering the constituent elements of the material in question and calculating the duration of exposure to the neutron flux and the energies of the incident neutrons. The licensee is responsible for performing direct measurements during actual removal and/or dismantlement of components. Those data will be used as the basis for specifying the necessary safety measures and procedures to maintain exposures at ALARA levels during the various dismantlement, removal, decontamination, and waste packaging and storage operations.

3.3.1.2.1 Pool Water The reactor pool and the primary cooling system contain approximately 50,000 gallons of water requiring removal. The water was supplied from potable water through filters and demineralizers. The cleanliness of the water was maintained by a system of filters, and H-OH demineralizers were used as necessary to maintain the conductivity to less than 5 micro-ohm per centimeter. Chemical additions to the water were not required to maintain the pH between 4.5 and 7.5. Table 3-2 lists the levels of radioactivity in the pool water measured in March 2004.

Table 3-2 Radioactivity of the Reactor Pool Water (March 17, 2004)

Gross Alpha <7.18 pCi/l Gross Beta 699 pCi/l Tritium 1,110,000 pCi/l Silver-108m 66.8 pCi/l Silver-110m 1,150 pCi/l Zinc-65 645 pCi/l 3.3.1.2.2 Bridge Suspension Frame and Grid Plate The reactor grid plate has overall dimensions of approximately 25 inches by 33 inches by 6 inches thick. Figure 3-2 illustrates the results of a survey of the reactor grid conducted by the licensee, using an Eberline RO-7 with a high-range probe in an underwater housing. The licensee believes a large contribution to the dose rates for the interior positions of the reactor grid, two rows in from each edge, originates from the 0.25 inch by 1 inch stainless steel alignment pins at these locations. Additionally, 18 stainless steel bolts (0.25 inch by 5 inches) are present around the outer edge of the reactor grid and were used to suspend the hopper from the bottom of the reactor grid.

Figure 3-2 Radiation Levels (R/h) on the Reactor Grid Plate (April 2004) 3.3.1.2.3 Heavy-Water Reflector According to the licensee, the last transfer of heavy water to the heavy-water reflector occurred in February 1992. Subsequently, heavy-water transfers to maintain the tritium content of the heavy-water reflector below 50 Ci were no longer required because of the removal of the 50-Ci limit (see Amendment 36 to the TSs). The licensee assumed that the reflector contained the maximum allowed activity of 50 Ci of tritium prior to the transfer and contained 44.6 Ci of tritium after completion of the 5-gallon transfer. The licensee calculated the annual tritium buildup in the heavy-water reflector using an average tritium production rate and the power history for each year and accounting for radioactive decay. The licensee evaluated the tritium inventory in the heavy-water reflector at the end of March 2004 to be 217 Ci. The heavy water is on loan from DOE and will be returned to the Savannah River Site.

3.3.1.2.4 Beam Ports The reactor staff have or will remove the collimators or experiment plugs that were installed in most of the 6-inch and 8-inch beam ports, with the following exceptions:

• The upper 6-inch through port, running east-west is believed to contain shielding materials and a 0.75-inch aluminum tube that runs through this shielding material and is open on the east and west ends of the through port. The contents of this through port will require detailed characterization before removal, processing, and disposal

• The beam port closest to the thermal column is believed to contain a collimator that extends the full length of the beam port (i.e., from the opening to the reactor core). This collimator is believed to contain a 0.5-inch stainless steel box in a taper, small at the reactor end and wide at the opening, around which lead and polyethylene shielding was attached. The contents of this beam port will require detailed characterization before removal, processing, and disposal. Dose rates of several R/h or higher are expected from the reactor end of this collimator. A dose rate of approximately 35 mrem per hour is present at the open end of the collimator, which extends from the beam port opening.

A 500-pound lead door or shutter shields each 6-inch beam port opening, and a 630-pound lead door shields each 8-inch beam port opening. These doors can be raised or lowered across the beam port opening (for the lower 6-inch through port, the door moves side to side rather than up and down).

The licensee stated that the contents of all remaining beam ports have been or will be removed and concrete port plugs reinstalled. Based upon its past experience, the licensee expects minimal contamination and little or no activation from these items.

3.3.1.2.5 Thermal Column The licensee recently opened the thermal column and compared it with the facility drawings.

This investigation by the licensee indicated that half of the graphite was removed in the late 1960s to early 1970s. Surveys of the graphite blocks showed elevated levels of contamination in areas where water had calcified, but radiation exposure levels were indistinguishable from background. Small samples of some of the graphite material were taken from the graphite blocks in the approximate center of the column, and results of the analysis of these samples are pending. From facility drawings, the thermal column contains the volumes of materials listed in Table 3-3 below.

The licensee stated that the extent the materials in the thermal column are surface contaminated or activated is not known. The licensee will perform detailed characterization before and after removal to determine the options for disposal.

Table 3-3 Estimated Material Volumes for the Thermal Column Material Volume (ft3) or Mass (lb)

Cadmium sheet 0.8 ft3 Boral (boron carbide between aluminum 1.0 ft3 cladding)

Graphite block 96 ft3 Lead shot 3,752 lb Lead block 16,747 lb Other lead (caulk and thin strip) 50 lb 3.3.1.2.6 Pneumatic Tube System Six of the eight tubes to the irradiation stations on the west side of the reactor grid are currently plugged at the point where the tubes penetrate the floor of the reactor pool. The licensee will perform detailed characterization before removal, processing, and disposal.

3.3.1.2.7 Other Items in the Reactor Pool The following are some of the other, higher level, radioactive components to be handled and processed during FNR decommissioning based on process knowledge and direct measurements performed by the licensee:

• reactor irradiation facility for large samples (RIFLS) reading as high as 50 R/hr

• heavy section steel irradiation (HSSI) experiment reading approximately 11,200 R/hr

• reactor control/shim rods reading about 2500 R/hr 3.3.1.3 Sanitary Sewer Lines According to the licensee, from the opening of the PML in 1954 until the summer of 1991, liquids containing low levels of radioactivity were discharged from the retention tanks in the PML to the sanitary sewer line following sampling to verify that applicable regulatory limits and license conditions were satisfied. The sanitary sewer line runs south along the western side of the PML, turns west and follows Bonisteel Boulevard towards the UM hospital, at which point it turns and runs along the river to the Ann Arbor sewage treatment plant. The licensee collected a sample of the internal pipe surface of the sewer line at the point it exits the PML. Attempts by the licensee to obtain sludge (solids) from several locations along this pathway were unsuccessful, given the small volume of sludge that was present at most locations. The samples, therefore, mainly consisted of liquid. There was no detectable radioactivity from the PML.

3.3.1.4 Soil beneath the Reactor Building Because of the possibility of pool water leakage and the 1993 loss of approximately 7500 gallons of low-level radioactive water, the licensee conducted an investigation to assess whether soil underlying the reactor pool around the reactor building was contaminated. During the investigation, the licensee performed soil borings (1) immediately north of the reactor pool through the first floor into the unexcavated area, (2) through the basement floor near the point where the foundation drain tile connects to the cold sump (the source of the 7500-gallon leak in 1993), and (3) immediately east of the drain-tile line just outside the reactor building.

This investigation detected only ambient levels of radionuclides normally present in soil, and tritium at a concentration of 14.5 pCi per gram in the upper foot of soil located immediately north of the reactor pool.

3.3.1.5 Ground Water Because of the 7500 gallons of low-level radioactive water released in 1993, and the possibility of leaks from the reactor pool itself, the licensee investigated the potential for radionuclides in the ground water near the reactor building. Since a previous monitoring well in this location was decommissioned because it dried up, the licensee established a new ground water monitoring well in April 2003 immediately south of the PML. Sampling of this well found, with the exception of 333 pCi/L of tritium (well below the U.S. Environmental Protection Agency maximum contaminant level of 20,000 pCi/L), no detectable radionuclides present in ground water other than those present in background water samples.

3.3.1.6 Radionuclides According to the licensee, radionuclides expected to be encountered during decommissioning of the FNR originated from reactor operations as well as experiments performed over the years.

Several of these radionuclides have short half-lives. The licensee determined the potential radionuclides present, shown in Table 3-4, through research of FNR historical documents and interviews with knowledgeable personnel.

During characterization of accessible areas, the licensee identified cobalt-60 and cesium-137 as the dominant contaminants with smaller amounts of numerous other activation and fission products. As such, the licensee concluded that the radionuclide mix does not appear to be uniform.

Table 3-4 List of Potential Radionuclides Half-Life Decay Nuclide Notes (yr) Mode Antimony-125 2.8 -, AP; from n-activation of materials containing tin Bismuth-210m 3.0x106 , AP; from n-activation of SS hardware Cadmium-109 1.26 , AP; from n-activation of cadmium metal or materials containing cadmium Carbon-14 5.73x103 - AP; from n-activation of graphite or materials containing carbon Cesium-134 2.1 -, AP; from n-activation of cesium, FP; minor FP inventory constituent Cesium-137 30.2 -, FP; expected to be predominant FP species present Cobalt-60 5.3 , -, +, AP; from n-activation of SS hardware; expected to be predominant AP species present Europium-152 13.5 -, AP/FP Europium-154 8.5 -, FP Iron-55 2.7 AP; from n-activation of SS hardware or materials containing iron Manganese-54 0.86 , AP; from n-activation of SS hardware Nickel-59 7.5x104 , AP; from n-activation of SS hardware Nickel-63 100 - AP; from n-activation of SS hardware Scandium-46 0.23 -, AP; from n-activation of materials used in testing/experiments Silver-108m 127 , AP; from n-activation of materials containing silver Silver-110m 0.68 -, AP; from n-activation of materials containing silver (Ag-110m)

Tritium 12.3 - AP; from n-activation of water and from shield tank Zinc-65 0.67 , +, AP; from n-activation of SS hardware

- = beta, + = positron, = electron capture, = gamma ray Note: The list of potential radionuclides provided above is based on the assumption that operations of the FNR have resulted in the neutron activation of reactor core components and other integral hardware or structural members that were situated adjacent to, or in close proximity to, the reactor core during operations. Specific items that are considered to have been exposed to neutron activation include materials composed of aluminum, steel, stainless steel, graphite, cadmium, lead, concrete, and possibly others. Neutron activation of materials beyond the concrete liner/biological shield structure (i.e., into surrounding soil volumes) is not expected for the FNR based on earlier studies, experience from similar research reactor decommissioning projects, reactor-specific calculations that considered measured values for neutron leakage fluence, integrated operating power histories, reactor core/pool structural configurations, and material composition of pool structures.

3.3.1.7 Conclusions The staff has reviewed the dose rates and contamination levels identified by the licensee and the licensees plans for followup surveys. Based on experience and professional judgment, the staff concludes that the licensees estimates of the radiological conditions and radiation measurements are acceptable. The staff finds that a followup characterization survey will be necessary following the removal of the material from the pool and pool draining. This survey will include the pool and any leakage pathways. Based on review of the information provided by the licensee, the staff concludes that the radiological status of the FNR has been adequately characterized and that this facility can be safely decommissioned.

3.3.2 Radiological Release Criteria The licensee proposed the DECON decommissioning alternative for the reactor. The licensee stated that the results of the site and facility radiological characterization survey indicate that the building structures may not need extensive decontamination to meet the release criteria.

The licensee proposed that the FSS will use derived concentration guideline levels (DCGLs) developed from the characterization survey data and the current NRC guidance for license termination in 10 CFR Part 20. The regulations in 10 CFR 20.1402 allow termination of a license and release of a site for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of a critical group of less than 25 mrem (0.25 millisievert) per year, and the residual radioactivity has been reduced to ALARA levels.

3.3.2.1 Structure Surfaces The licensee proposed that for remediation activities, it will select the DCGLs for residual radioactive material contamination on FNR structural surfaces from the tables of NRC default screening values (refer to NUREG-1757, Consolidated NMSS Decommissioning Guidance).

Table 3-5 lists the screening values for total structure surface contamination; guideline levels for removable activity are 10 percent of the values in the table. The NRC has conservatively evaluated these default screening levels as satisfying the goal that doses to facility occupants and the public during future facility use do not exceed 25 mrem annually. Default screening criteria are based on conservative exposure scenario and pathway parameters and are generally regarded as providing a high level of confidence that the annual dose limits will not be exceeded. These screening values are applicable where it can be demonstrated that the residual radioactivity is present on the surface only and volumetric contamination (less than 10 millimeters deep) is not present.

Table 3-5 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces Acceptable Screening Radionuclide Symbol Levels1,2 for Unrestricted Release (dpm/100 cm2)3 3

Tritium H 1.2E+08 14 Carbon-14 C 3.7E+06 22 Sodium-22 Na 905E+03 35 Sulfur-35 S 1.3E+07 36 Chlorine-36 Cl 5.0E+05 54 Manganese-54 Mn 3.2E+04 55 Iron-55 Fe 4.5E+06 60 Cobalt-60 Co 7.1E+03 63 Nickel-63 Ni 1.8E+06 90 Strontium-90 Sr 8.7E+03 90 Technetium-99 Tc 1.3E+06 129 Iodine-129 I 3.5E+04 137 Cesium-137 Cs 2.8E+04 192 Iridium-192 Ir 7.4E+04 Notes:

1 Screening levels presented here are taken from the NRCs Supplemental Information on the Implementation of the Final Rule on Radiological Criteria for License Termination, issued 1998.

The DP states that the licensee will develop site-specific screening levels for the project in the manner described in that reference.

2 Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases in which the fraction of removable contamination is undetermined or higher than 0.1, users may assume for screening purposes that 100 percent of the surface contamination is removable, and therefore the screening levels should be decreased by a factor of

10. Users may calculate site-specific levels based on available data on the fraction of removable contamination and DandD version 2.

3 Units are dpm/100 cm2; 1 dpm is equivalent to 0.0167 becquerel (Bq). Therefore, to convert to units of Bq/square meter (m2), multiply each value by 1.67. The screening values represent surface concentrations of individual radionuclides that would be deemed in compliance with the 0.25 millisievert per year (mSv/yr) (25 mrem/yr) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the sum of fractions rule applies (see Note 4 in Appendix B to 10 CFR Part 20).

Characterization surveys performed by the licensee have identified multiple radionuclide contaminants on surfaces and in various media at the FNR. Predominant contaminants anticipated by the licensee at the time of license termination are cobalt-60 and cesium-137.

However, additional fission and activation products are present on some surfaces, generally at lower concentrations and at spotty distributions. The licensee described that, for surfaces, it will determine concentrations of specific contaminants and ratios to their respective DCGLs to demonstrate satisfaction of the Unity Rule as described in Section 4.3.3 of the MARSSIM (Ref. 9). The licensee will use gross beta measurements to demonstrate compliance with surface activity guidelines, and it will base the gross beta DCGL on measurements of surrogate contaminants with known relationships to the total contamination mix.

The DCGLs described are net (above background) concentrations and activity levels of radionuclides; the licensee will make appropriate adjustments for instrument background levels and naturally occurring radionuclide concentrations in various media before comparing data to the respective DCGLs.

Because of the conservatism used in the development of the default screening values, further evaluations and actions are not required to reduce residual radioactivity to ALARA levels.

3.3.2.2 Surface Soil and Sediment The licensee proposed that for remediation activities, it will select the DCGLs for residual radioactive material contamination in sediments or surface soil (top 15 cm of soil) under or near the FNR from the tables of NRC default screening values (refer to NUREG-1757). Table 3-6 lists the screening values for contaminants in soil. These default screening levels provide assurance that doses to facility occupants and the public during future facility use do not exceed 25 mrem annually. These default screening criteria are based on conservative exposure scenario and pathway parameters and are generally regarded as providing a high level of confidence that the annual dose limits will not be exceeded.

Table 3-6 Acceptable License Termination Screening Values of Common Radionuclides for Surface Soil Radionuclide Symbol Surface Screening Values1,2 3

Tritium H 1.1E+02 14 Carbon-14 C 1.2E+01 22 Sodium-22 Na 4.3E+00 35 Sulfur-35 S 2.7E+02 36 Chlorine-36 Cl 3.6E-01 45 Calcium-45 Ca 5.7E+01 46 Scandium-46 Sc 1.5E+01 54 Manganese-54 Mn 1.5E+01 55 Iron-55 Fe 1.0E+04 57 Cobalt-57 Co 1.5E+02 60 Cobalt-60 Co 3.8E+00 59 Nickel-59 Ni 5.5E+03 63 Nickel-63 Ni 5.5E+03 90 Strontium-90 Sr 1.7E+00 94 Niobium-94 Nb 5.8E+00 99 Technetium-99 Tc 1.9E+01 129 Iodine-129 I 5.0E-01 134 Cesium-134 Cs 5.7E+00 137 Cesium-137 Cs 1.1E +/- 01 152 Europium-152 Eu 8.7E +/- 00 154 Europium-154 Eu 8.0E +/- 00 192 Iridium-192 Ir 4.1E +/- 01 210 Lead-210 Pb 9.0E-01 226 Radium-226 Ra 7.0E-01 Radionuclide Symbol Surface Screening Values1,2 Radium-226+C3 226 Ra+C 6.0E-01 227 Actinium-227 Ac 5.0E-01 227 Actinium-227+C Ac+C 5.0E-01 228 Thorium-228 Th 4.7E+00 228 Thorium-228+C Th+C 4.7E+00 230 Thorium-230 Th 1.8E+00 230 Thorium-230+C Th+C 6.0E-01 232 Thorium-232 Th 1.1E+00 232 Thorium-232+C Th+C 1.1E+00 231 Protactinium-231 Pa 3.0E-01 231 Protactinium-231+C Pa+C 3.0E-01 234 Uranium-234 U 1.3E+01 235 Uranium-235 U 8.0E+00 235 Uranium-235+C U+C 2.9E-01 238 Uranium-238 U 1.4E+01 238 Uranium-238+C U+C 5.0E-01 238 Plutonium-238 Pu 2.5E+00 239 Plutonium-239 Pu 2.3E+00 241 Plutonium-241 Pu 7.2E+01 241 Americium-241 Am 2.1E+00 242 Curium-242 Cm 1.6E+02 243 Curium-243 Cm 3.2E+00 Notes:

1 These values represent surface soil concentrations of individual radionuclides that would be deemed in compliance with the 0.25 mSv/yr (25 mrem/yr) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the sum of fractions rule applies (see Note 4 in Appendix B to 10 CFR Part 20).

2 Screening values are in units of pCi/g equivalent to 0.25 mSv/yr (25 mrem/yr). To convert from pCi/g to units of Bq per kilogram (Bq/kg), divide each value by 0.027. These values were derived using DandD screening methodology (NUREG/CR-5512, Volume 3, Residual Radioactive Contamination for Decommissioning). They were derived based on selection of the 90th percentile of the output dose distribution for each specific radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at Standard Man or at the mean of the distribution for an average human.

3 Plus Chain (+C) indicates a value for a radionuclide with its decay progeny present in equilibrium.

The values are concentrations of the parent radionuclide but account for contributions from the complete chain of progeny in equilibrium with the parent radionuclide (NUREG/CR-5512, Volumes 1, 2, and 3).

Characterization surveys performed by the licensee have identified multiple radionuclide contaminants at the FNR that could also be present in soil and sediment. Predominant contaminants anticipated by the licensee are cobalt-60 and cesium-137. However, additional fission and activation products could also be present in soil and sediment. In addition, the licensee described that variable radionuclide mixtures may be present in soil and sediment.

Therefore, the licensee will determine concentrations of specific significant contaminants and ratios to their respective DCGLs in a manner satisfying the Unity Rule, as described in Section 4.3.3 of the MARSSIM (Ref. 9).

The criteria described are net (above background) concentrations and activity levels of radionuclides; the licensee will make appropriate adjustments for instrument background levels and naturally occurring radionuclide concentrations in various media before comparing data to the respective criteria.

Because of the conservatism used in establishing the default screening values, further evaluations and actions to demonstrate that the final conditions satisfy ALARA provisions are not required.

3.3.2.3 Subsurface and Inaccessible Structures The criteria for residual radioactive contamination on FNR facility surfaces discussed in Section 3.3.2.1 are not applicable for surfaces where the contaminant is not at the surface (greater than 10 millimeters deep), activated surfaces, and inaccessible areas excluding buried pipes, etc.

The licensee must still develop the criteria for radioactive contamination of these types of surfaces because it has not yet obtained characterization results. However, it will develop the specific release criteria that will be applied in these instances at a later date using RESRAD-BUILD or equivalent methodology. The licensee will develop the criteria to ensure that estimated doses to facility occupants and the public during future facility use is less than 25 mrem annually.

Characterization surveys performed by the licensee have identified multiple radionuclide contaminants on surfaces and in various media at the FNR. Predominant contaminants anticipated by the licensee at the time of proposed license termination are cobalt-60 and cesium-137. However, additional fission and activation products are present in some media, generally at lower concentrations and at spotty distributions. The licensee described that variable radionuclide mixtures are also present for different media. The licensee will determine concentrations of specific significant contaminants and ratios to their respective DCGLs in a manner satisfying the Unity Rule, as described in Section 4.3.3 of the MARSSIM (Ref. 9).

The License Termination Rule (10 CFR Part 20, Subpart E, Radiological Criteria for License Termination) also requires that residual radioactivity resulting from licensed material for release to unrestricted use must be at ALARA levels. The licensee may need to further reduce the criteria for residual radioactive material contamination of subsurface structures or components within the physical structure of the FNR facility (left after remediation or decontamination) to satisfy the ALARA requirement. Reduction of the cleanup criteria for subsurface and inaccessible structures may follow an examination by the licensee of the reduction in the estimated dose to the facility occupants and the public using the RESRAD-BUILD software combined with an examination of the costs associated with achieving these reduced levels of residual radioactivity. The licensee will document this evaluation in its final report to the NRC.

The criteria described in this section should be net (above background) concentrations and activity levels of radionuclides; the licensee will make appropriate adjustments for instrument background levels and naturally occurring radionuclide concentrations in various media before comparing data to the respective criteria.

3.3.2.4 Conclusions The licensee has adequately specified the radiological release criteria need for license termination that will be used for accessible building surfaces and soil. The staff concludes that the licensee understands the release criteria for license termination for the FNR and has proposed acceptable DCGLs in accordance with applicable guidance.

3.3.3 Decommissioning Tasks 3.3.3.1 Characterization Surveys The licensee has conducted characterization studies as part of the planning activities for the DP. The licensee has identified the type, quantity, condition, and location of radioactive and/or hazardous materials that are or may be present in the FNR. It conducted extensive surveys of accessible areas of the FNR in September 2002 and April 2003. The characterization report provided in Appendix A to the DP summarizes the results of these surveys. The licensee will perform additional surveys in conjunction with the dismantlement and decontamination activities discussed below, as previously inaccessible areas are made accessible.

3.3.3.2 Dismantlement and Decontamination of the Facility Dismantling and decontamination will be required to remove materials that were activated or radiologically contaminated during operation of the FNR in order to meet the unconditional release criteria for license termination. The licensee will employ standard industry dismantling and decontamination techniques using tools such as wire saws, high-pressure/ultra high-pressure water, needle guns, jack hammers, torches/plasma arc torches, hydraulic cutters, and hand tools, following approved procedures or work packages. The following sections discuss typical dismantling and decontamination activities. The licensee may opt not to follow the sequence for ALARA, safety, accessibility, or scheduling reasons.

3.3.3.2.1 Systems Formerly Important to Safety As decommissioning progresses, the licensee may inactivate (deenergize and isolate) or remove all inactive systems or systems not currently required by the TSs or later decommissioning activities but formerly identified in the SAR. The licensee has identified several systems, structures, or components that will be removed from the facility in accordance with the change control process defined in 10 CFR 50.59 and Section 9.0 of the DP, including the following:

• standby generator

• heavy-water reflector

• spent fuel storage racks

• pneumatic tube system external to the reactor pool

• secondary cooling system

• emergency cooling system

• control console

• exhaust for hood in Room 3103

• exhaust for pneumatic blowers, first-floor trunks around pool, and storage ports

• beam port extensions 3.3.3.2.2 Other Systems Systems identified by the licensee that may be deenergized and/or isolated include the potable water line, drain lines to the hot or cold sump, reactor air to miscellaneous supplies, gaseous nitrogen supply lines, and the demineralized water supply to the PML. Systems interfacing with the contiguous wall of the PML will be isolated on the PML side of the interface, when practical.

The licensee will apply the quality assurance requirements identified in Section 1.3.4.2 of the DP when required.

3.3.3.2.3 Asbestos The licensee will remove, package, and dispose of radioactively contaminated asbestos-containing materials in accordance with applicable regulations. It may also remove, survey, and dispose of uncontaminated asbestos-containing materials in accordance with applicable regulations.

3.3.3.2.4 Temporary Systems The licensee may need to install temporary systems to support decommissioning activities.

These may include additional electrical outlets for temporary ventilation or decontamination equipment, a water purification system to purify or decontaminate liquids, openings in the reactor building for equipment access or waste removal, waste storage and handling systems or equipment, service air, potable waste, fire detection, and fire hose stations.

3.3.3.2.5 Reactor Pool The licensee will estimate the radioactivity associated with the high-dose items (the reactor grid, shim and control rods, beam port extensions, etc.) when these items become accessible.

The licensee will reduce the size of the reactor grid plate, shim and control rods, heavy-water reflector, pneumatic tubes, RIFLS, HSSI experiment, and remaining miscellaneous high-dose items to facilitate loading into HICs or, for inherently stable items, a liner. To do this, the licensee may use long-reach tools, remotely operated equipment, human divers, or a combination of these techniques.

The licensee plans to use the water in the reactor pool as shielding and for contamination control during high-dose item size reduction and removal activities. However, it may be necessary for the licensee to lower the water level or drain the pool to remove items such as the pneumatic tube bundle penetration that could, upon removal, introduce a potential pool drainage pathway. If the pool water levels are lowered, the licensee may use shielding or remote size reduction techniques to maintain personnel exposure at ALARA levels.

The licensee may transfer high dose-rate items, such as the shim and control rods, RIFLS, and HSSI experiment, to the hot cells in the PML for size reduction. High dose-rate items may also be transferred and loaded dry into the HIC or liner using shields.

Once the high dose-rate items are loaded into the HIC or liner, the licensee will place the HIC or liner into an approved, shielded shipping cask for transport to an approved disposal site. The HIC or liner should be directly loaded into a shipping cask submerged in the reactor pool (similar to the methods when loading and shipping irradiated fuel elements), whenever the size of the cask permits. The licensee recognizes that the HIC or liner may require indirect loading using a shielded transfer cask if the size of the cask or other factors prohibits loading in the reactor pool.

The licensee will dispose of the water in the reactor pool when the water is no longer useful as a radiological shield or for contamination control. The licensee will filter and treat the liquid from the pool and piping as necessary to meet discharge requirements of the license as well as Federal, State, and local laws. Liquid effluents will subsequently be discharged to the City of Ann Arbor sanitary sewer using approved procedures. The licensee will treat, stabilize, and package liquids not meeting release criteria to meet shipping requirements and waste acceptance criteria at an approved disposal site.

Following draining of the pool, the licensee will characterize the structure to determine the extent and depth of activation and contamination in the reactor pool floor, walls, and embedded beam port tubes. UM may use the characterization results to select either the pool removal or pool decontamination option for decommissioning based on ALARA, safety, structural, cost, schedule, and future use considerations.

Future licensee plans for the reactor building require the decontamination and removal of the reactor pool from the building. UM has elected to remove those portions of the reactor pool that may not be readily remediated. Contingent upon the results of the reactor pool characterization, the reactor pool walls and possibly portions of the reactor pool floor will be cut into large blocks and packaged and shipped as radioactive waste by the licensee to a licensed disposal facility. The licensee may not remove materials embedded in the concrete (beam port tubes, drain pipes, conduit, tile, etc.) unless it is necessary to meet transportation requirements and the disposal site waste acceptance criteria.

If decontamination of the reactor pool or a portion of the reactor pool is elected for the decommissioning option, then the licensee will decontaminate pool surfaces and the activated concrete will be removed to levels that will facilitate termination of the license. The licensee will collect core bore samples to evaluate subsurface contamination. The licensee stated that contamination present below surfaces (e.g., surface cracks or voids) will be decontaminated or removed, and the waste generated will be packaged and shipped to a licensed disposal site.

3.3.3.2.6 Embedded Pipes The licensee will decontaminate or remove contaminated pipes, drains, and conduit embedded in concrete. Sludge, scale, and other waste generated will be treated or stabilized and packaged to meet the disposal site waste acceptance criteria. The licensee will discharge decontamination liquid to the sanitary sewer if it meets the license requirements as well as Federal, State, and local requirements for discharge to the sewer.

3.3.3.2.7 Surface and Subsurface Sampling The licensee proposed to collect sufficient soil samples from unexcavated areas beneath and west of the pool to determine if an unknown leak in the pool contaminated the soil surrounding the pool. The licensee will seal or plug any holes drilled through the concrete to prevent the hole from becoming a potential pathway to the environment.

3.3.3.2.8 Contaminated Equipment The licensee will remove or decontaminate contaminated equipment from each floor of the FNR. Essential equipment, such as heating, ventilation, and air conditioning (HVAC) and electrical and instruments interfacing with PML or FNR systems, may be isolated to reduce the potential for accidental releases of water or energy. Examples of equipment that may need to be decontaminated or removed include the following:

• basementprimary coolant piping and instrumentation, holdup tank, primary pump and motor, ion exchange piping and system, and hot and cold sump pumps and motors

• first floorHVAC ducts, source storage ports, transfer chute, thermal column and thermal column door, and drain lines and piping not embedded in concrete

• second floorHVAC equipment, ducts and butterfly valves

• third floorreactor bridge, remaining reactor suspension frame, pool and reactor instrumentation, heavy-water reflector support equipment, HVAC, drain lines and piping, pool filter/vacuum system, and any miscellaneous low-dose items in or attached to the pool

• fourth floorcrane over the pool, HVAC (contamination not expected in all components) 3.3.3.2.9 Remaining Areas The licensee will decontaminate or remove any remaining contaminated areas within the FNR, then survey to confirm the area has been decontaminated to levels that will meet unconditional release criteria. Examples of areas that may require decontamination include the following:

• basementconcrete floor, hot and cold sumps, holdup tank pit, ion exchange pit, and walls

• first floorfloor, wall by the source storage ports, and thermal column door trench

• third floorlaboratories, floor around the pool, and the south wall The licensee does not expect decontamination to be required on the second and fourth floors.

The licensee will package and dispose of waste generated during this activity at a licensed disposal site.

3.3.3.2.10 Soil and Buried Pipe Remediation If contaminated soil is identified and the source of the contamination is the FNR, the licensee will evaluate the results against the release criteria. If contamination levels require soil removal, the licensee will remove, package, and dispose of the soil at an approved disposal site.

The licensee will package and dispose of any buried pipes (e.g., drain tiles) found to be radiologically contaminated that cannot be decontaminated on site to meet final release criteria.

The licensee will collect final release samples after remediation of the soil or buried pipes.

However, the excavations will remain open to permit the NRC to perform confirmatory surveys or sampling.

The licensee will collect split samples before backfilling if backfilling is necessary for safety reasons before confirmatory surveys are performed. The licensee will notify the NRC of the expected completion date of the remediation so that the NRC has the opportunity to be present to verify collection of soil samples. Once NRC concurrence is received, the licensee will backfill the excavation to reduce any potential safety hazard.

The assumption that neutron activation of the soil beneath the reactor pool did not occur will be confirmed by evaluating the activation of the concrete floor in the void directly beneath the reactor core, which is accessible from the reactor basement (refer to Figures 2-2 and 2-7).

3.3.3.3 Final Survey and Report Following decontamination and remediation activities of the FNR, the licensee will perform a final radiological survey covering the entire FNR. A final radiological survey, executed according to the approved FSS plan, will document that the licensees decommissioning efforts achieve the release criteria.

Once all decontamination has been performed and verified through final radiological surveys, the licensee will develop a final release report. The licensee will record in this report the decontamination and remediation activities performed and document the final radiological status of the FNR facility and associated grounds. The licensee will use this final report in part as the basis of the application for license termination.

3.3.3.4 Conclusions Based on review of the information provided by the licensee, the plans for decommissioning the FNR facility follow an acceptable sequence and are acceptable to the NRC staff.

3.3.4 Schedule The scheduled time from regulatory approval of the DP to the request for release of the site for unrestricted use is estimated to be 15 months. The licensee proposed that changes to the schedule may be made at the discretion of UM, including changes due to resource allocation, availability of a radioactive waste burial site, interference with ongoing UM activities, ALARA considerations, further characterization measurements, and/or temporary onsite radioactive waste storage operations.

3.3.4.1 Conclusions Based on a review of the licensees proposed decommissioning schedule, the staff concludes that the licensees proposed schedule is acceptable.

3.3.5 Proposed Final Status Survey Plan The licensee provided a plan for the development, review, and approval of the FSS plan once the site is fully characterized. The licensees stated objective of the FSS is to ensure that the facility meets the unrestricted release criteria.

3.3.5.1 General Survey Approach The licensee noted that all factors influencing the FSS for the FNR are not available and will not be available until it evaluates more facility details following additional characterization activities to be conducted upon approval of the DP. The outline for the proposed FSS plan prepared by the licensee is intended to provide information to the NRC for determining the adequacy of the licensees understanding of the proposed FSS plan as it pertains to the goal of remediation in a manner satisfying the radiological criteria for license termination. The final FSS plan, which the licensee will formally submit for approval at a later date (included as a license condition; refer to Section 4.0), will adequately demonstrate compliance with the radiological criteria for license termination.

The licensee prepared its proposed FSS plan in accordance with the guidelines and recommendations presented in the MARSSIM (Ref. 9). The licensee committed to implement the MARSSIM process that emphasizes the use of data quality objectives (DQOs) and data quality assessment, along with a quality assurance and quality control program. As such, the licensee will follow the graded approach concept of the MARSSIM to assure that survey efforts are maximized in those areas having the greatest potential for residual contamination.

The licensee committed to conducting the FSS with trained radiological control technicians, who follow standard, written procedures and use properly calibrated instruments, sensitive to the potential contaminants.

The licensee may develop designs for specific surveys for some areas, including determination of specific nuclide mixture guidelines, sampling or measurement methods, survey unit identification and classification, and data evaluation techniques, at the time of the survey in accordance with the guidance presented in the proposed FSS plan.

3.3.5.2 Instrumentation The licensee will base the selection of instruments on the type of radiation emitted for the radionuclides of interest, as well as the required range, accuracy, and tolerance needed to demonstrate conformance to specified requirements. Selection and use of instrumentation for the FSS will also be based upon the need to ensure that the residual radioactivity remaining on site meets the release criteria. Table 3-7 lists the instrumentation the licensee intends to use for the FSS and associated documentation (e.g., characterization information used in the design of the final survey), along with estimated detection sensitivities. The licensee will also accept other instruments that are the functional equivalent of those listed.

Table 3-7 Instrumentation for FNR Radiological Surveys Sensitivity (dpm/100 cm2, except as noted)

Detector Type Make Meter Application Static Count Scanning (1 minute) 43-68 Gas Ludlum 2221 Gross beta scan and 1200 500 Proportional measurement 43-68 Gas Ludlum 2221 Nickel-53 Gross beta 5000 2000 Proportional scan and measurement 43-67 Floor Ludlum 2221 Gross beta scan 800 N/A Monitor 43-68 Gas Ludlum 2221 Gross alpha 200 70 Proportional measurement Tennelec Gas Tennelec N/A Gross alpha smear N/A 5 LB5100 Proportional measurement Tennelec Gas Tennelec N/A Gross beta smear N/A 10 LB5100 Proportional measurement 44-10 NaI Ludlum 2221 Gamma scan 10 pCi/g N/A Because the radionuclides expected by the licensee to be present as contaminants emit (with few exceptions) beta particles with maximum energies greater than 0.300 megaelectron volts (MeV), detector efficiencies for measuring surface activity are generally determined using technetium-99 (maximum beta energy of approximately 0.292 MeV). For situations in which contaminants emit beta particles of lower energy (e.g., facilities contaminated with nickel-63),

the licensee will specifically determine detector efficiencies for those contaminants.

The licensee will account for the effects of surface conditions on surface activity measurements through the use of a source efficiency factor, in accordance with the guidance in ISO-7503-1, Evaluation of Surface Contamination, Part 1, Beta Emitters and Alpha Emitters, issued August 1998 (Ref. 13), and NUREG/CR-1507 (Ref. 10). The licensee general considers default source efficiency factors of 0.5 for beta emitters greater than 0.4 MeV maximum energy and 0.25 for beta emitters between 0.150 MeV and 0.400 MeV maximum to be applicable to anticipated FNR contaminants and surface conditions. However, if contaminants or conditions are not consistent with use of these default values, the licensee will determine specific source efficiency factors and document them in the FSS design.

The licensee will estimate detection sensitivities using the guidance in the MARSSIM (Ref. 9) and NUREG/CR-1507 (Ref. 10). The licensee will choose instrumentation and survey techniques with the objective of achieving detection sensitivities of 25 percent of the criteria for structure surfaces, for both scanning and direct measurement, to ensure identification of areas of elevated activity having a size and activity level that could adversely impact the average residual activity level for the survey units.

The licensee will follow guidance from equipment manufacturers and the American National Standards Institute (ANSI) N323-1978, American National Standard Radiation Protection Instrumentation Test and Calibration, issued 1978 (Ref. 14), for calibration methods, calibration interval, and operational and background quality control checks. The licensee will establish procedures to implement this guidance and will perform instrument calibrations using standards traceable to NIST or an equivalent standards organization.

3.3.5.3 Data Quality Objectives The licensee designed its stated DQOs to achieve a 95-percent confidence level that the release criteria are met. The survey design will be based on both Type I () and Type II ()

decision errors of 5 percent. The DP describes data quality indicators for precision, accuracy, representativeness, completeness, and comparability as follows:

• Precision is determined by comparison of replicate values from field measurements and sample analyses; the objective is a relative percent difference of 20 percent or less at 50 percent of the release criteria.

• Accuracy is the degree of agreement with the true or known value; the objective for this parameter is +/-20 percent at 50 percent of the release criteria.

• Representativeness and comparability do not have numeric values. Performance is assured through selection and proper implementation of sampling and measurement techniques.

• Completeness refers to the portion of the data that meets acceptance criteria and is thus acceptable for statistical testing; the objective for this parameter is 90 percent.

3.3.5.4 Classifications of Areas by Contamination Potential For FNR areas determined to be impacted areas per guidance in the MARSSIM, the licensee adopted the following definitions that describe three classifications of areas, according to contamination potential.

(1) Class 1 areas are impacted areas that, prior to remediation, are expected to have concentrations of residual radioactivity that exceed the guideline value.

(2) Class 2 areas are impacted areas that, prior to remediation, are not expected to have concentrations of residual radioactivity that exceed the guideline value.

(3) Class 3 areas are impacted areas that have a low probability of containing residual activity.

Typically levels will not exceed 25 to 35 percent of the guideline value.

The licensee used facility history, including the Historical Site Assessment, issued 2003 (Ref.

15), and radiological monitoring conducted during characterization and remedial activities as the bases for classification. Once the licensee obtains approval for the FSS plan through a subsequent license amendment request to the NRC, the licensee may make changes to the classification of an area as long as the classification is changed to one of higher contamination potential. The licensee will obtain a license amendment pursuant to 10 CFR 50.90, Application for Amendment of License or Construction Permit, if the change would decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3), as discussed in Section 4.0.

3.3.5.5 Identification of Survey Units A survey unit is a portion of a facility with common contaminants and contamination potential and contiguous surfaces or areas. The licensee will identify survey units following remediation, at the time of FSS design. Table 4-3 of the DP provides a listing of facility areas that are currently expected to be included in the FSS, the estimated surface areas, anticipated contamination potential classifications, and the projected number of survey units within each area. The licensee developed this listing based on the historical assessment, preliminary survey data obtained in November 2002, and the characterization survey performed in April 2003. The DP notes that the licensee will determine actual survey unit boundaries and classifications at the time of FSS design, and survey unit classifications and surface areas may change as characterization and remedial activities proceed. If classifications and boundaries change, the licensee will redesign the FSS for the affected areas and reevaluate data as necessary.

3.3.5.6 Demonstrating Compliance The null hypothesis recommended for use in the MARSSIM and selected by the licensee is stated, The residual radioactivity in the survey unit exceeds the release criterion. Rejection of the null hypothesis by the statistical test therefore concludes that the residual activity does not exceed guidelines and the survey unit satisfies requirements for unrestricted release.

The licensee will use nonparametric statistical tests recommended in the MARSSIM to demonstrate that radiological conditions satisfy the established criteria. One of the tests is the Wilcoxon Rank Sum (WRS) test. The licensee may use the WRS test when a specific radionuclide of concern is present in background at a concentration greater than 10 percent of the guideline level and when the measurement is not radionuclide specific (e.g., for direct measurements of total surface activity). The licensee may use the Sign test when the radionuclide of concern is not present in background at a significant fraction (i.e., less than 10 percent) of the guideline level. The Sign test will also be used when evaluating data based on the Unity Rule and may be used for surface activity data representing multiple surface media. Both of these tests are applicable to the FNR facility FSS, and the licensee will not be able to evaluate FSS data using statistical tests without first obtaining NRC approval. The licensee will select a specific test method when designing the FSS.

3.3.5.7 Background Reference Areas and Materials The licensee will determine background contributions if (1) the residual contamination includes a radionuclide that occurs in background or (2) measurements are not radionuclide specific.

The licensee anticipates that the FSS will require multiple reference areas and materials. For applications involving the WRS test, reference areas will be of the same material as the survey unit being evaluated, but without a history of potential contamination by licensed operations.

The licensee will obtain a set of reference measurements for each instrument used for survey unit evaluation. For applications involving the Sign test, sufficient background determinations will be made for each media or surface material and with each instrument to provide an average background level that is accurate to within +/-20 percent (usually requires a minimum of 8 to 10 measurements). The licensee will identify reference area and background requirements at the time of individual survey unit FSS design.

3.3.5.8 Final Status Survey Design 3.3.5.8.1 Sample Size and Sampling Locations The licensee provided adequate information that will be used for determining the data needs for the statistical tests for each survey unit. The licensee indicated that the FSS design for that survey unit will document the following information:

• calculation of the relative shift (/)

• / = DCGL - lower bound of gray region

• DCGL, as the gross or nuclide-specific release criteria

• lower bound of the gray region initially selected as half of the DCGL as recommended by the MARSSIM

• determined empirically from actual survey data; however, for planning purposes, equals a value of 25 percent of the DCGL

• decision errors established by DQOs for this project of 0.05 for both Type I and Type II errors

• determination of the number of data points required as obtained from MARSSIM Tables 5.3 (WRS test) and 5.5 (Sign test)

The MARSSIM recommends a triangular measurement or sampling pattern to increase the probability of identifying small areas of residual activity. The licensee will use this type of triangular pattern for the FSS, except where dimensions and/or other factors related to a specific survey unit require use of an alternate pattern. If the systematic pattern does not provide sufficient data points to satisfy the number determined as outlined above, the licensee will locate additional data points using a random-number technique.

3.3.5.8.2 Scan Surveys Licensee data collected in support of the FSS of structure surfaces will consist of scans to identify locations of residual contamination, direct measurements of beta surface activity, and measurements of removable beta surface activity. The FSS data collected by the licensee for open land (soil) areas will consist of scans to identify locations of residual contamination and samples of soil, analyzed for potential contaminants. The licensee will obtain additional measurements and samples as necessary to supplement the information from these typical survey activities.

The licensee will use gas-flow proportional detectors for beta surface scans. Floor monitors with 580-cm2 detectors will be used for floor and other larger accessible horizontal surfaces; hand-held 125-cm2 detectors will be used for surfaces not assessable with the floor monitor.

When scanning, (1) the detector will be within 0.5 cm of the surface (if surface conditions prevent this distance, the detection sensitivity for an alternate distance will be determined and the scanning technique adjusted accordingly), (2) scanning speed will be no greater than one detector width per second, and (3) audible signals will be monitored and locations of elevated direct levels identified for further investigation. The licensee committed to the minimum scan coverages of 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have the highest potential for contamination.

The licensee will use NaI gamma scintillation detectors (2 inch x 2 inch) for gamma surface scans of structures and open land areas to identify locations of residual surface activity. When scanning, (1) the detector will be moved in a serpentine pattern, while advancing at a rate of approximately 0.5 meters per second, (2) the distance between the detector and the surface will be maintained within 5 centimeters of the surface, and (3) audible signals will be monitored and locations of elevated direct levels identified for further investigation. The licensee committed to the minimum scan coverages of 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased toward areas considered by professional judgment to have the highest potential for contamination.

3.3.5.8.3 Direct Measurements and Sampling The licensee will perform direct measurement of beta surface activity at designated locations using a 125-cm2 gas-flow detector. Measurements will be conducted by integrating the count over a 1-minute period. Where adverse surface conditions may result in underestimating activity by direct measurements, the licensee will obtain surface samples for laboratory analyses. The licensees FSS design will identify the need for such sampling for specific survey units.

The licensee will collect a smear sample for removable activity at each direct surface activity measurement location with a 2-inch diameter cloth or paper filter by wiping a 100-cm2 surface area using moderate pressure. Dampened smears will be used to sample for removable tritium activity.

The licensee will obtain samples of surface (upper 15 centimeters) soil from selected locations using a hand trowel or bucket auger. Approximately 500 to 1000 grams of soil will be collected at each sampling location.

3.3.5.9 Data Assessment The licensee committed to review radiological data needed to support the FSS to assure that the type, quantity, and quality are consistent with the survey plan and design assumptions.

Data standard deviations will be compared with the assumptions made in establishing the number of data points.

The licensee will compare individual and average data values with guideline values and confirm proper survey area classifications.

The licensee will investigate individual measurement data in excess of the guideline level for Class 2 areas and in excess of 25 percent of the guideline for Class 3 areas. Anomalies and deviations from design assumption and plan requirements will be identified. Need for investigation, reclassification, remediation, and/or resurvey will be determined. In addition, the licensee will initiate a corrective action will be initiated, as appropriate, and repeat the data conversion and assessment process for new data sets.

3.3.5.10 Final Status Survey Report The licensee will prepare an FSS report describing the survey procedures and findings for submission to the NRC in support of license termination. The FSS report will provide a complete record of the facilitys radiological status and a comparison to the site release criteria. The licensees FSS report will provide a summary of any ALARA analysis, survey data results, and overall conclusions, which collectively demonstrate that the FNR facility meets the radiological criteria for unrestricted use. The FSS report will include information such as the number and type of measurements, basic statistical quantities, and statistical test results. It will also contain additional detail to enable an independent or third party re-creation and evaluation of the survey results and a determination as to whether the site release criteria have been met.

The following outline from Section 4.15 of the DP illustrates a general format that the licensee may use for the FSS report:

• a summary of the results of the FSS

• a discussion of any changes in the FSS process that were proposed in the license termination plans or other prior submittals

• a description of the method used to determine the number of samples for each survey unit

• a summary of the assumed parameters used to calculate the number of samples and a justification for these values The FSS report will also provide the results for each survey unit, including the following:

• the number of survey samples collected for the survey unit

• a map or drawing of the survey unit showing the reference system and random start systematic sample locations for Class 1 and 2 survey units, and random locations shown for Class 3 survey units and reference areas

• measured sample concentrations

• statistical evaluation of the measured concentrations

• judgmental and miscellaneous sample data sets reported separately from those samples collected for performing the statistical evaluation

• discussion of anomalous data including any areas of elevated direct radiation detected during scanning that exceeded the investigation level or measurement locations in excess of the DCGL

• a statement that a given survey unit satisfied the DCGL and the elevated measurement comparison if any sample points exceeded the DCGL

• a description of any changes in initial survey unit assumptions relative to the extent of residual radioactivity

• a description of the investigation conducted when the data from a survey unit fail to ascertain the reason for the failure and a discussion of the impact that the failure has on the conclusion that the facility was ready for final radiological surveys

• a description of the impact a survey unit failure has on other survey unit information and the reason for the failure The licensee may adjust this outline to more clearly present the information. The level of detail will be sufficient to clearly describe the FSS program and certify the results.

3.3.5.11 Change Control The staff reviewed the DP to determine whether it listed sufficient criteria to establish the types of changes to equipment, structures, system components, and procedures that would be permissible without prior NRC approval. The staff, recognizing that change control criteria needed to support reactor operations were not well suited for determining the types of changes expected to occur during decommissioning, issued Amendment No. 47 for the FNR on January 29, 2004, to maintain the authority to make changes to the facility and procedures without prior Commission approval as contained in 10 CFR 50.59. Therefore, the following change control criteria that would support changes that may be needed to implement the FSS during decommissioning would not require prior NRC approval:

• The licensee may make changes to the DP without prior approval provided the proposed changes do not require Commission approval pursuant to 10 CFR 50.59 use a statistical test other than the Sign test or WRS test for evaluation of the FSS increase the radioactivity level, relative to the applicable DCGL, at which an investigation occurs reduce the coverage requirements for scan measurements decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3) increase the Type I decision error result in more than a minimal increase in the environmental consequences not previously evaluated in the final SAR (as updated) foreclose the release of the site for possible unrestricted use

• The licensee shall submit reports of any characterization surveys performed that were not part of the license amendment application and shall submit the completed FSS plan for review prior to performing the FSS.

The staff finds that the change control criteria proposed by the licensee will adequately facilitate changes needed to implement the FSS in a manner that ensures both the safety of workers and the public and facilitates timely decommissioning of the FNR.

3.3.5.12 Conclusions The staff has reviewed the licensees DP concerning the planning of the FSS. The staff finds that the licensee has adequate experience to develop and implement an acceptable MARSSIM FSS. Once the licensee develops the FSS plan, it will present the plan for review and approval prior to implementation. The NRC staff concludes this aspect of the DP meets the requirements of 10 CFR 50.82(b)(4)(iii) and is therefore acceptable.

3.4 Estimated Cost The licensee stated that decommissioning of the FNR will be accomplished without dismantlement of the building. Table 1-1 of the DP presents the detailed estimated cost to decommission the FNR licensed areas. The factors used in these cost estimates were based upon a detailed cost estimate. Using the High cost in Table 1-1 of the DP, the licensee estimated that the project will cost up to $9,781,173. Based on the given High estimate, the DP states that UM is committed to providing funding for decommissioning of the FNR, in accordance with 10 CFR 50.75(e)(iv).

3.4.1 Conclusions The staff has reviewed the licensees decommissioning cost estimate and finds that the cost estimates are consistent with the scope of work covering decommissioning of the FNR. The licensee stated that the UM Regents have specifically approved the expenditure of funds from investment proceeds sufficient to cover the High cost estimate. The staff concludes that UM is committed to providing acceptable funding for decommissioning of the FNR.

3.5 Quality Assurance 3.5.1 Overview Section 1.3.4.1 of the DP briefly describes the quality assurance programs used during decommissioning, summarized as follows:

• A quality assurance program is applied to the design, fabrication, construction, and testing of structures, systems, and components of the facility. These quality assurance requirements would apply to the remediation activities conducted.

• A quality assurance program, which may or may not be the same as the above-mentioned program, is applied to the design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing operations, maintenance, repair, and modification of components of packaging used in the transportation of licensed material.

• Additional quality assurance requirements are applied to the FSS and associated documentation (e.g., characterization information used in the design of the FSS) to ensure that data and the analysis of the data provided to the NRC in the FSS report are accurate and complete.

3.5.2 Quality Assurance for Design, Construction, Testing, Modification, and Maintenance The FNR has a quality assurance program, as discussed in Section 1.3.4.2 of the DP, that meets the requirement in 10 CFR 50.34, Contents of Applications; Technical Information, for establishing and executing a quality assurance program for the design, construction, testing, modification, and maintenance of a research reactor. The descriptions of the managerial and administrative controls will result in a revision to the current quality assurance program. The FNR will continue to maintain this quality assurance program for the design, construction, testing, modification, and maintenance (including remediation activities) of the reactor.

UM will continue to require that all contractors and subcontractors participating in design, construction, testing, modification, and maintenance (including remediation) activities follow the established quality assurance program. Contractors and subcontractors may recommend or request changes to the quality assurance program. UM may or may not make changes to the quality assurance program after review against applicable guidance or standards recommended.

Changes to the quality assurance program will be approved as discussed in Section 2.4 of the DP.

3.5.3 Quality Assurance for Packaging, Preparation for Shipment, and Transportation of Licensed Material Subpart H, Quality Assurance, of 10 CFR Part 71 specifies the requirements for packaging, preparation for shipment, and transportation of licensed material. The managerial and administrative controls the FNR has established to satisfy the requirements of this subpart, described in Section 2.4 of the DP, differ slightly from those previously used. The NRC has approved the current FNR quality assurance program as required by 10 CFR 71.101(c). The licensee will follow the existing quality assurance program and maintain it through timely renewal, as necessary, to support packaging, preparation for shipment, and transportation of licensed material during remediation activities.

UM will continue to require that all contractors and subcontractors participating in packaging, preparation for shipment, and transportation of licensed material follow the approved quality assurance program. Contractors and subcontractors may recommend or request changes to the quality assurance program. UM may or may not make changes to the quality assurance program after review against the requirements of 10 CFR Part 71, Subpart H. The licensee will submit revisions to the quality assurance program to the NRC for approval as required by 10 CFR 71.101(c) prior to implementation and use for the packaging, preparation for shipment, and transportation of licensed materials.

UM may elect to use a contractors or subcontractors quality assurance program to fulfill the requirements contained in 10 CFR Part 71, Subpart H, after verification that the contractors or subcontractors quality assurance program is acceptable to UM and has been approved by the NRC.

3.5.4 Quality Assurance for Final Status Survey and Associated Documentation 3.5.4.1 General UM is responsible for developing a FSS quality assurance program and associated documentation (e.g., characterization information used in the design of the FSS). This program will be reviewed and approved as described in Section 2.4 of the DP. The FSS quality assurance program will incorporate the appropriate regulatory requirements applicable to the planning and conduct of radiological surveys necessary for the termination of the FNR license and the release of the site for unrestricted use. The quality assurance program implements the appropriate criteria in Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. The following sections describe the required components of the FSS quality assurance program.

3.5.4.2 Organization Section 2.4 of the DP identifies the written definitions of authorities, duties, and responsibilities of managerial, operations, and safety personnel; a defined organizational structure; assigned responsibility for review and approval of plans, specifications, designs, procedures, data, and reports; and assigned responsibility for procurement and oversight of services (e.g., analytical laboratory). The licensee will give personnel assigned organizational responsibility for performing quality assurance functions the necessary independence and authority to allow them to identify quality problems; to initiate, recommend, and provide solutions; and to verify implementation of solutions.

The reactor manager has the authority and responsibility for implementing all aspects of the FSS quality assurance program. The reactor manager will ensure that survey activities meet the requirements outlined in the FSS quality assurance program to safeguard the decommissioning staff, the UM community, and the public. The reactor manager will regularly review the adequacy of the FSS quality assurance program and provide an assessment to the Director and the review committee. The reactor manager will inform the appropriate UM decommissioning staff and contractors of decommissioning activities related to the FSS quality assurance program.

The project manager will ensure that the contractor complies with the FSS quality assurance program and satisfies the objectives and requirements for FSS. Furthermore, the project manager is responsible for ensuring that all activities are performed in a manner to permit the termination of the FNR license and the release of the site for unrestricted use. In accordance with the American Society of Mechanical Engineers Quality Assurance Requirements for Nuclear Facility Applications, issued 2001 (Ref. 16), the individual(s) or organization(s) responsible for establishing and executing the FSS quality assurance program may delegate any or all of the work to others but will otherwise retain responsibility.

3.5.4.3 Written Quality Assurance Program The licensee will establish a documented quality assurance program for the FSS and associated documentation (e.g., characterization information used in the design of the FSS) at the earliest practical time, consistent with the schedule for accomplishing the activities. The licensee will document this quality assurance program through written polices, procedures, or instructions and will execute it through the conduct of FSS activities and creation of associated documentation in accordance with those policies, procedures, or instructions. Activities for the FSS and creation of associated documentation affecting quality will be accomplished under suitably controlled conditions. Controlled conditions included the use of appropriate equipment, suitable environmental conditions for accomplishing the activity, and assurance that prerequisites for the given activity have been satisfied. The quality assurance program will provide for any special controls, processes, survey equipment, tools, and skills to attain the required quality of activities and items and for verification of that quality.

3.5.4.4 Training Personnel will be qualified for their assigned duties before working independently or will be under the direct supervision of a qualified individual. Personnel performing special processes will be qualified according to specific codes and standards or in accordance with national consensus documents. Qualification will include proficiency demonstrated by each individual, both initially and then periodically. Qualification will also be demonstrated when required by the designated codes or standards.

The licensee will maintain training records that include the trainees name, dates of training, types of training, test results, protective equipment use authorizations, and instructors names.

Care will be taken to ensure that properly qualified instructors conduct all training. As the primary criterion, persons responsible for the presentation of training should have knowledge and experience in the process or subject matter. It is desirable that trainers also have the presentation skills or classroom conduct appropriate to the level of the training being presented.

For those with limited background in training, early instruction should be monitored and feedback should be provided.

3.5.4.5 Quality Assurance Records The licensee will ensure that sufficient records are specified, prepared, reviewed, authenticated, and maintained to reflect the achievement of the required quality. Records will include documents such as operating logs, results of reviews, inspections, tests, assessments, work performance monitoring, and material or sample analyses. Records will be identifiable, available, and retrievable. The records will be reviewed to ensure their completeness and ability to serve their intended function. Requirements will be established concerning record collection, safekeeping, retention, maintenance, updating, location, storage, preservation, administration, and assigned responsibility. Requirements will be consistent with applicable regulations, as well as the potential for impact on quality and radiation exposure to workers and the public.

The licensee will identify documents that require control, including policies, procedures, or instructions that specify quality requirements or describe activities affecting quality, such as instructions, procedures, and drawings. Qualified personnel will review policies, procedures, or instructions (including revisions) for conformance with technical requirements and quality system requirements and approve them as discussed in Section 2.4 of the DP. The personnel performing relevant activities will ensure the currency of policies, procedures, or instructions requiring control. The licensee will take measures to ensure that personnel understand the document controls to be used. Obsolete or superseded documents will be identified and measures will be taken to prevent their use.

The licensee will control all documents related to the FSS using appropriate policies, procedures, or instructions. All significant changes to such documents will be similarly controlled. This documentation normally would include a survey plan, survey packages, survey results, and a survey report.

3.5.4.6 Control of Measuring Equipment Measures will be established to assure that instruments and other measuring devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.

The licensee will base the selection of instruments on the type of radiation emitted for the radionuclides of interest, as well as the required range, accuracy, and tolerance needed to demonstrate conformance to specified requirements. Selection and use of instrumentation for the FSS will also be based upon the need to ensure that the residual radioactivity remaining on site meets the release criteria. Table 3-7 lists the instrumentation intended for use for the FSS and associated documentation (e.g., characterization information used in the design of the final survey), along with estimated detection sensitivities. Other instruments, which are the functional equivalent of those listed, will also be acceptable.

Calibration procedures will identify or reference required accuracy. Methods of evaluating the accuracy of instrumentation will be defined in procedures and will follow ANSI N323-1978 (11).

The calibration method and interval of calibration for instruments will be defined, based on the type of equipment, stability characteristics, required accuracy, intended use, manufacturers recommendations, and other conditions affecting capability, and will follow ANSI N323-1978.

Out-of-calibration and defective instruments will be removed from service and not used until they have been repaired and recalibrated. The licensee will repair or replace any instruments consistently found to be out of calibration.

Measuring instruments will be calibrated at prescribed time periods or immediately before use and whenever the accuracy of the equipment is suspect. Calibration will be performed using standards traceable to NIST or an equivalent standard organization. Instruments found to be out of calibration will require a documented evaluation, commensurate with the significance of the condition, of the validity of data obtained with that instrument since its previous acceptable performance. Instruments will be properly handled and stored to maintain accuracy according to ANSI N323-1978. The licensee will suitably mark or otherwise identify instruments to indicate calibration status.

Operational and background checks will be performed at the beginning of each day of FSS activity and whenever there is reason to question instrument performance. These checks should follow ANSI N323-1978.

3.5.4.7 Audits and Corrective Actions Project audits will be planned and conducted using criteria that describe acceptable work practices, including performance. Audits will verify compliance with applicable requirements of the FSS quality assurance program and will determine its effectiveness. The scheduling of audits and allocation of resources will be based on the work status, risk, and complexity of the item or process being assessed. Audits will be performed and results reported as described in Section 2.4 of the DP. Conditions adverse to quality will be identified to the reactor manager promptly and corrected as soon as practicable. Significant conditions adverse to quality will be identified to the licensees review committee as soon as practicable, along with the cause of the condition, when known, and corrective actions taken to prevent recurrence.

3.5.4.8 Conclusions The currently approved quality assurance program will be in place for activities leading up to the FSS, such as remediation and transportation of licensed material. The staff has reasonable assurance that an adequate quality assurance plan is in place and implemented in accordance with 10 CFR 50.82(b)(4)(v) for these activities.

The information presented in the DP provides reasonable assurance to the staff that a FSS quality assurance program constructed according to the stated requirements will adequately address the necessary quality functions associated with decommissioning activities in accordance with 10 CFR 50.82(b)(4)(v).

3.6 Physical Security The regulations in 10 CFR 73.67(c)(1) require facilities to maintain a physical security plan when they possess special nuclear materials of moderate strategic significance or 10 kilograms or more of special nuclear material of low strategic significance. Because all special nuclear material in the form of reactor fuel covered by the license for the FNR has been removed, and the license has been amended for no possession of reactor fuel (Amendment No. 47), a physical security plan is not required.

It is recognized that the regulations in 10 CFR Part 20, Subpart I, Storage and Control of Licensed Material, apply to the remaining byproduct and special nuclear materials possessed by the FNR. All FNR licensed materials that are in storage will be secured from unauthorized access or removal, and licensed materials that are not in storage will be under the control and constant surveillance of authorized FNR personnel as required by 10 CFR Part 20.

3.6.1 Conclusions Based on the NRC staffs review, the licensee has acceptable security access controls to prevent inadvertent exposure to workers and members of the public.

4.0 ADDITIONAL LICENSE CONDITIONS The regulations in 10 CFR 50.82(b)(5) state in part that the licensees DP will be approved by license amendment subject to such conditions and limitations as the NRC deems appropriate and necessary. Based on the requirements of the regulations and the staffs review of the licensees application, the staff has added the following conditions to the UM FNR license:

The license is amended to approve the decommissioning plan described in the licensees application dated June 23, 2004, as supplemented on January 05, 2006, and authorizes inclusion of the decommissioning plan as a supplement to the Safety Analysis Report pursuant to 10 CFR 50.82(b)(5).

A license amendment pursuant to 10 CFR 50.59 shall be obtained for changes to this decommissioning plan if the change would:

• Require Commission approval pursuant to 10 CFR 50.59;

• Use a statistical test other than the Sign test or Wilcoxon Rank Sum test for evaluation of the final status survey;

• Increase the radioactivity level, relative to the applicable derived concentration guideline level, at which an investigation occurs;

• Reduce the coverage requirements for scan measurements;

• Decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3);

• Increase the Type I decision error;

• Result in more than a minimal increase in the environmental consequences not previously evaluated in the final safety analysis report (as updated);

• Foreclose the release of the site for possible unrestricted use.

The licensee shall submit reports of any characterization surveys performed that are not part of the license amendment application and shall submit the completed final status survey plan for review prior to performing the final status survey.

The above license conditions make the licensees DP part of the Safety Analysis Report for the facility in accordance with the regulations, help to ensure that changes to the DP that may impact compliance with the release criteria in the regulations in Part 20 are not made without NRC review, and ensure that important information to the decommissioning process still under development by the licensee are submitted to the NRC when complete.

4.1 Conclusions The staff has added requirements to the UM FNR license in accordance with the regulations in 10 CFR 50.82(b)(5). The staff concludes that these license conditions are necessary to meet the requirements of 10 CFR 50.82(b)(5) and to allow the licensee to develop the final radiological survey and documentation necessary to permit the staff to make the required findings to terminate the license in accordance with 10 CFR 50.82(b)(6).

5.0 TECHNICAL SPECIFICATIONS The licensees organization for decommissioning is changing substantially. To support these changes, the licensee proposed revisions to TS 6.0, Administrative Controls. Those changes were issued with Amendment No. 49 to Facility License No. R-28 (Ref. 17). The NRC will issue Amendment No. 49 concurrently with the decommissioning amendment approving the UM FNR DP.

5.1 Conclusions With the issuance of Amendment No. 49 to Facility License No. R-28 for the UM FNR reactor, appropriate changes have been made to support the UM FNR DP and the safe decommissioning of the reactor.

6.0 ENVIRONMENTAL CONSIDERATION

The Commission has prepared an EA and Finding of No Significant Impact, published in the FR on February 6, 2006 (71 FR 6104-6105). On the basis of the EA and this safety evaluation, the Commission has determined that no environmental impact statement is required and that issuance of this license amendment approving decommissioning will have no significant adverse effect on the quality of the human environment.

7.0 CONCLUSION

S Based on the staffs review of the licensees application for approval of decommissioning, the staff finds that the licensee is adequately cognizant of its continuing responsibilities to protect the health and safety of both workers and the public from undue radiological risk. The DP provides reasonable evidence that the licensee is prepared to dismantle the reactor and dispose of all significant reactor-related radioactive materials in accordance with applicable regulations and applicable NRC guidance.

The staff concludes that the choice of the DECON decommissioning alternative is acceptable and meets the requirements of 10 CFR 50.82(b)(4)(i) for decommissioning without significant delay.

The staff concludes that the DP provides acceptable organizational structure and control to decontaminate and dismantle the FNR while maintaining due regard for protecting the public, environment, and workers from significant radiological risk. Furthermore, the staff concludes that the licensees plan for radiation protection and radioactive material and waste management is acceptable based on the use of standard guidance and practices for such programs. The staff finds the personnel training program that FNR proposed in the DP to be acceptable because its scope covers all aspects of decommissioning activities that need to be performed safely. The industrial safety program and procedural and equipment controls are consistent with such programs at decommissioning reactors and are therefore acceptable. The staff concludes that potential radiological consequences attributable to the types of accidents that could occur during decommissioning are well within acceptable limits. The staff concludes that the licensees DP contains a description of the controls and limits on procedures and equipment to protect occupational and public health and safety as required by 10 CFR 50.82(b)(4)(ii).

The staff concludes that the licensee has adequately described the radiological status of the FNR facility and has proposed acceptable release criteria for the FNR facility. The licensee has acceptably described the tasks, sequence of activities, and schedule needed to decommission the FNR facility. The staff also concludes that the licensee has provided an acceptable description of its planned final radiation survey as required by 10 CFR 50.82(b)(4)(iii).

The staff concludes that the licensee has provided, in accordance with 10 CFR 50.82(b)(4)(iv),

an acceptable updated cost estimate for the DECON decommissioning alternative and has an acceptable plan for assuring the availability of adequate funds for the completion of decommissioning.

The licensee has provided a description of TSs, quality assurance provisions, and physical security plan provisions to be in place during decommissioning. The staff has determined that these aspects of the DP meet the regulations in 10 CFR 50.82(b)(4)(v). Therefore, based on the discussion above, the staff concludes that the licensees DP meets the requirements of 10 CFR 50.82 (b)(4).

The staff has concluded, on the basis of the considerations discussed above, that (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from any accident previously evaluated, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities, and (3) such activities will be conducted in compliance with the Commissions regulations and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public.

ABBREVIATIONS AEC airborne effluent concentration ALARA as low as reasonably achievable ANSI American National Standards Institute AP activation product Bq becquerel CFR Code of Federal Regulations Ci curie(s) cm centimeter(s) cm2 square centimeter(s)

Cs cesium DAW dry active waste DCGL derived concentration guideline level DECON decontamination decommissioning option DOE U.S. Department of Energy DP decommissioning plan dpm disintegration(s) per minute DQO data quality objective FNR Ford Nuclear Reactor FP fission product FR Federal Register FSS final status survey ft3 cubic foot/feet g gram(s) h hour(s)

HEPA high-efficiency particulate air HIC high-integrity containers HSSI heavy section steel irradiation HVAC heating, ventilation, and air conditioning in. inch(es)

JHA job hazard analysis kg kilogram(s)

L liter(s) lb pound(s)

LSA low specific activity m2 square meter(s)

MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual mCi millicurie(s)

MeV megaelectron Volts MIOSHA Michigan Occupational Safety and Health Act of 1974 mL milliliter mrem millirem mSv millisievert Mwt megawatt thermal NaI sodium iodide NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission pCi picocurie(s)

PML Phoenix Memorial Laboratory PPE personal protective equipment R roentgen RIFLS reactor irradiation facility for large samples RSO radiation safety officer RWP radiation work permit SAR safety analysis report SS stainless steel TEDE total effective dose equivalent (see 10 CFR Part 20)

TS technical specification UM University of Michigan WRS Wilcoxon Rank Sum yr year REFERENCES

1. University of Michigan, Ford Nuclear ReactorAmendment, Decommissioning Plan, June 18, 2004.

2. University of Michigan, Decommissioning Plan for the Ford Nuclear Reactor, Rev. 00, June 23, 2004.

3. University of Michigan, Decommissioning Plan for the Ford Nuclear Reactor, Rev. 01, January 5, 2006.

4. University of Michigan, Additional Detail on the In Situ Gamma Spectroscopy Data, January 10, 2006.

5. U.S. Nuclear Regulatory Commission, NUREG-1700, Rev. 1, Standard Review Plan for Evaluating Nuclear Power License Termination Plans, Appendix 2, Washington DC, April 2003.

6. U.S. Nuclear Regulatory Commission, NUREG-1537, Rev. 0, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Appendix 17.1, Washington, DC, February 1996.

7. U.S. Nuclear Regulatory Commission, Circular No. 81-07, Control of Radioactively Contaminated Material, Washington, DC, May 14, 1981.

8. U.S. Nuclear Regulatory Commission, Information Notice No. 85-92, Surveys of Wastes Before Disposal from Nuclear Reactor Facilities, Washington, DC, December 2, 1985.

9. U.S. Nuclear Regulatory Commission, NUREG-1575, Rev. 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), Washington, DC, August 2000.

10. U.S. Nuclear Regulatory Commission, NUREG/CR-1507, Final, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, Washington, DC, June 1997.

11. U.S. Nuclear Regulatory Commission, NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, Washington, DC, 1988.

12. U.S. Nuclear Regulatory Commission, Regulatory Guide 8.15, Rev. 1, Acceptable Programs for Respiratory Protection, Washington, DC, October 1999.

13. International Organization for Standardization, Evaluation of Surface ContaminationPart 1: Beta Emitters and Alpha Emitters (First Edition), ISO-7503-1, August 1998.

14. American National Standards Institute, ANSI N323-1978, American National Standard Radiation Protection Instrumentation Test and Calibration, 1978.

15. CH2M HILL, Inc., Historical Site Assessment, Ford Nuclear Reactor, North Campus, University of Michigan, Final Report, Richland, Washington, 2003.

16. American Society of Mechanical Engineers, Quality Assurance Requirements for Nuclear Facility Applications, New York, NY, 2001.

17. U.S. Nuclear Regulatory Commission, Amendment No. 49 to Facility Operating License No. R-28 University of Michigan Ford Nuclear Reactor, Appendix A, May , 2006.