ML013520290

Issuance of Amendment Elimination of Requirements for Post-Accident Sampling

ML013520290
Person / Time
Site:	Millstone
Issue date:	01/08/2002
From:	Nerses V NRC/NRR/DLPM/LPD1
To:	Price J Dominion Nuclear Connecticut
NersesV, NRR/DLPM, 415-1484
References
TAC MB2721
Download: ML013520290 (14)

Text

January 8, 2002 Mr. J. A. Price Vice President - Nuclear Technical Services - Millstone Dominion Nuclear Connecticut, Inc.

c/o Mr. David A. Smith Rope Ferry Road Waterford, CT 06385

SUBJECT:

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 - ISSUANCE OF AMENDMENT RE: ELIMINATION OF REQUIREMENTS FOR POST-ACCIDENT SAMPLING (TAC NO. MB2721)

Dear Mr. Price:

The Commission has issued the enclosed Amendment No. 201 to Facility Operating License No. NPF-49 for the Millstone Nuclear Power Station, Unit No. 3, in response to your application dated July 31, 2001.

The amendment deletes Technical Specifications Section 6.8.4.d, Post-Accident Sampling," for Millstone Nuclear Power Station, Unit No. 3 and thereby eliminates the requirements to have and maintain the post-accident sampling program.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Victor Nerses, Sr. Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 201 to NPF-49

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

January 8, 2002 Mr. J. A. Price Vice President - Nuclear Technical Services - Millstone Dominion Nuclear Connecticut, Inc.

c/o Mr. David A. Smith Rope Ferry Road Waterford, CT 06385

SUBJECT:

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 - ISSUANCE OF AMENDMENT RE: ELIMINATION OF REQUIREMENTS FOR POST-ACCIDENT SAMPLING (TAC NO. MB2721)

Dear Mr. Price:

The Commission has issued the enclosed Amendment No. 201 to Facility Operating License No. NPF-49 for the Millstone Nuclear Power Station, Unit No. 3, in response to your application dated July 31, 2001.

The amendment deletes Technical Specifications Section 6.8.4.d, Post-Accident Sampling," for Millstone Nuclear Power Station, Unit No. 3 and thereby eliminates the requirements to have and maintain the post-accident sampling program.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Victor Nerses, Sr. Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 201 to NPF-49

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

PUBLIC OGC PDI-2 R/F ACRS EAdensam TClark CCowgill, RI WReckley GHill (2) WBeckner JClifford VNerses Accession Number: ML013520290 OFFICE CLIIP/LPM* PDI-2/PM PDI-2/LA PDI-2/SC NAME WReckley JHarrison for SLittle for REnnis for VNerses TClark JClifford DATE 9-12-01 1-3-02 1-2-02 1-8-02 OFFICIAL RECORD COPY

Millstone Nuclear Power Station Unit 3 cc:

Ms. L. M. Cuoco Mr. Evan W. Woollacott Senior Nuclear Counsel Co-Chair Dominion Nuclear Connecticut, Inc. Nuclear Energy Advisory Council Rope Ferry Road 128 Terry's Plain Road Waterford, CT 06385 Simsbury, CT 06070 Edward L. Wilds, Jr., Ph.D. Mr. D. A. Christian Director, Division of Radiation Senior Vice President - Nuclear Operations Department of Environmental Protection and Chief Nuclear Officer 79 Elm Street Innsbrook Technical Center - 2SW Hartford, CT 06106-5127 5000 Dominion Boulevard Waterford, CT 06385 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. C. J. Schwarz 475 Allendale Road Master Process Owner - Operate the Asset King of Prussia, PA 19406 Dominion Nuclear Connecticut, Inc.

Rope Ferry Road First Selectmen Waterford, CT 06385 Town of Waterford 15 Rope Ferry Road Senior Resident Inspector Waterford, CT 06385 Millstone Nuclear Power Station c/o U.S. Nuclear Regulatory Commission Mr. P. J. Parulis P. O. Box 513 Process Owner - Oversight Niantic, CT 06357 Dominion Nuclear Connecticut, Inc.

Rope Ferry Road Mr. G. D. Hicks Waterford, CT 06385 Master Process Owner - Training Dominion Nuclear Connecticut, Inc.

Mr. W. R. Matthews Rope Ferry Road Vice President and Senior Waterford, CT 06385 Nuclear Executive - Millstone Dominion Nuclear Connecticut, Inc. Mr. R. P. Necci Rope Ferry Road Vice President - Nuclear Operations - Millstone Waterford, CT 06385 Dominion Nuclear Connecticut, Inc.

Rope Ferry Road Ernest C. Hadley, Esquire Waterford, CT 06385 P.O. Box 1104 West Falmouth, MA 02574-1104 Mr. John Markowicz Co-Chair Nuclear Energy Advisory Council 9 Susan Terrace Waterford, CT 06385

Millstone Nuclear Power Station Unit 3 cc:

Mr. D. A. Smith Process Owner - Regulatory Affairs Dominion Nuclear Connecticut, Inc.

Rope Ferry Road Waterford, CT 06385 Ms. Nancy Burton 147 Cross Highway Redding Ridge, CT 00870 Mr. William D. Meinert Nuclear Engineer Massachusetts Municipal Wholesale Electric Company P.O. Box 426 Ludlow, MA 01056

DOMINION NUCLEAR CONNECTICUT, INC., ET AL.

DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 201 License No. NPF-49

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

A. The application for amendment by the applicant dated July 31, 2001, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate 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 Commission's regulations; 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; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 201, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. Dominion Nuclear Connecticut, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA REnnis for/

James W. Clifford, Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: January 8, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 201 FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Appendix A Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 6-16 6-16

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 201 TO FACILITY OPERATING LICENSE NO. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 DOCKET NO. 50-423

1.0 INTRODUCTION

By letter dated July 31, 2001, the Dominion Nuclear Connecticut, Inc., (the licensee), submitted a request for changes to the Millstone Nuclear Power Station, Unit No. 3 Technical Specifications (TSs). The requested changes would delete TS Section 6.8.4.d, Post-Accident Sampling," and thereby eliminate the requirements to have and maintain the post-accident sampling program (PASS).

In the aftermath of the accident at Three Mile Island (TMI), Unit 2, the U.S. Nuclear Regulatory Commission (NRC) imposed requirements on licensees for commercial nuclear power plants to install and maintain the capability to obtain and analyze post-accident samples of the reactor coolant and containment atmosphere. The desired capabilities of the PASS were described in NUREG-0737, Clarification of TMI Action Plan Requirements. The NRC issued orders to licensees with plants operating at the time of the TMI accident to confirm the installation of PASS capabilities (generally as they had been described in NUREG-0737). A requirement for PASS and related administrative controls was added to the TSs of the operating plants and was included in the initial TSs for plants licensed during the 1980s and 1990s. Additional expectations regarding PASS capabilities were included in Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident.

Significant improvements have been achieved since the TMI accident in the areas of understanding risks associated with nuclear plant operations and developing better strategies for managing the response to potentially severe accidents at nuclear power plants. Recent insights about plant risks and alternate severe accident assessment tools have led the NRC staff to conclude that some TMI Action Plan items can be revised without reducing the ability of licensees to respond to severe accidents. The NRCs efforts to oversee the risks associated with nuclear technology more effectively and to reduce unnecessary regulatory burdens on licensees have prompted the NRC to consider eliminating the requirements for PASS in the TSs and other parts of the licensing bases of operating reactors.

The staff has completed its review of the topical reports submitted by the Combustion Engineering Owners Group (CEOG) and the Westinghouse Owners Group (WOG) that proposed the

elimination of PASS. The justifications for the proposed elimination of PASS requirements center on evaluations of the various radiological and chemical sampling and their potential usefulness in responding to a severe reactor accident or making decisions regarding actions to protect the public from possible releases of radioactive materials. As explained in more detail in the staffs SEs for the two topical reports, the staff has reviewed the available sources of information for use by decision-makers in developing protective action recommendations and assessing core damage.

Based on this review, the staff found that the information provided by PASS is either unnecessary or is effectively provided by other indications of process parameters or measurement of radiation levels. The staff agrees, therefore, with the owners groups that licensees can remove the TS requirements for PASS, revise (as necessary) other elements of the licensing bases, and pursue possible design changes to alter or remove existing PASS equipment.

2.0 BACKGROUND

In a letter dated October 26, 1998, as supplemented by letters dated April 28, 1999, April 10 and May 22, 2000, the WOG submitted the Topical Report WCAP-14986, Post Accident Sampling System Requirements: A Technical Basis. The report provided evaluations of the information obtained from PASS samples to determine the contribution of the information to plant safety and accident recovery. The report considered the progression and consequences of core damage accidents and assessed the accident progression with respect to plant abnormal and emergency operating procedures, severe accident management guidance, and emergency plans. The report provided the owners group's technical justification for the elimination for the various PASS sampling requirements. The specific samples and the staffs findings are described in the following evaluation.

The NRC staff prepared a model SE relating to the elimination of requirements on post accident sampling and solicited public comment (65 FR 49271) in accordance with the consolidated line item improvement process (CLIIP). The use of the CLIIP in this matter is intended to help the NRC to efficiently process amendments that propose to remove the PASS requirements from the TSs.

Licensees of nuclear power reactors to which this model apply were informed (65 FR 65018) that they could request amendments confirming the applicability of the SE to their reactors and providing the requested plant-specific verifications and commitments.

3.0 EVALUATION The technical evaluation for the elimination of PASS sampling requirements for Westinghouse nuclear steam supply system plants is provided in the SE dated June 14, 2000, for the WOG Topical Report WCAP-14986. The NRC staffs SE approving the topical report is located in the NRCs Agencywide Documents Access and Management System (ADAMS) (Accession Number ML003723268).

The ways in which the requirements and recommendations for PASS were incorporated into the licensing bases of commercial nuclear power plants varied as a function of when plants were licensed. Plants that were operating at the time of the TMI accident are likely to have been the subject of confirmatory orders that imposed the PASS functions described in NUREG-0737 as obligations. The issuance of plant-specific amendments to adopt this change, which would remove PASS and related administrative controls from the TSs, supersede the PASS specific requirements imposed by post-TMI confirmatory orders.

As described in its SE for the topical report, the staff finds that the following PASS sampling requirements may be eliminated for plants of Westinghouse designs:

1. reactor coolant dissolved gases

2. reactor coolant hydrogen

3. reactor coolant oxygen

4. reactor coolant pH

5. reactor coolant chlorides

6. reactor coolant boron

7. reactor coolant conductivity

8. reactor coolant radionuclides

9. containment atmosphere hydrogen concentration

10. containment oxygen

11. containment atmosphere radionuclides

12. containment sump pH

13. containment sump chlorides

14. containment sump boron

15. containment sump radionuclides The staff agrees that sampling of radionuclides is not required to support emergency response decision making during the initial phases of an accident because the information provided by PASS is either unnecessary or is effectively provided by other indications of process parameters or measurement of radiation levels. Therefore, it is not necessary to have dedicated equipment to obtain this sample in a prompt manner.

The staff does, however, believe that there could be significant benefits to having information about the radionuclides existing post-accident in order to address public concerns and plan for long-term recovery operations. As stated in the SE for the topical report, the staff has found that licensees could satisfy this function by developing contingency plans to describe existing sampling capabilities and what actions (e.g., assembling temporary shielding) may be necessary to obtain and analyze highly radioactive samples from the reactor coolant system (RCS), containment sump, and containment atmosphere. (See item 4.1 under Verifications and Commitments.) These contingency plans must be available to be used by a licensee during an accident; however, these contingency plans do not have to be carried out in emergency plan drills or exercises. The contingency plans for obtaining samples from the RCS, containment sump, and containment atmosphere may also enable a licensee to derive information on parameters such as hydrogen concentrations in containment and boron concentration and pH of water in the containment sump.

The staff considers the sampling of the containment sump to be potentially useful in confirming calculations of pH and boron concentrations and confirming that potentially unaccounted for acid sources have been sufficiently neutralized. The use of the contingency plans for obtaining samples would depend on the plant conditions and the need for information by the decision-makers responsible for responding to the accident.

In addition, the staff considers radionuclide sampling information to be useful in classifying certain types of events (such as a reactivity excursion or mechanical damage) that could cause fuel damage without having an indication of overheating on core exit thermocouples. However, the staff agrees with the topical report's contentions that other indicators of failed fuel, such as letdown radiation monitors (or normal sampling system), can be correlated to the degree of failed fuel.

(See item 4.2 under Verifications and Commitments.)

In lieu of the information that would have been obtained from PASS, the staff believes that licensees should maintain or develop the capability to monitor radioactive iodines that have been released to offsite environs. Although this capability may not be needed to support the immediate protective action recommendations during an accident, the information would be useful for decision-makers trying to limit the publics ingestion of radioactive materials. (See item 4.3 under Verifications and Commitments.)

The staff believes that the changes related to the elimination of PASS that are described in the topical report, related SE and this proposed change to the TSs are unlikely to result in a decrease in the effectiveness of the licensees emergency plan. The licensee, however, must evaluate possible changes to its emergency plan in accordance with 10 CFR 50.54(q) to determine if the change decreases the effectiveness of its site-specific plan. The licensee should perform the appropriate evaluations and report changes to its emergency plan in accordance with applicable regulations and procedures.

The staff notes that redundant, safety-grade, containment hydrogen concentration monitors are required by 10 CFR 50.44(b)(1), are addressed in NUREG-0737 Item II.F.1 and Regulatory Guide 1.97, and are relied upon to meet the data reporting requirements of 10 CFR Part 50, Appendix E, Section VI.2.a.(i)(4). The staff concludes that during the early phases of an accident, the safety-grade hydrogen monitors provide an adequate capability for monitoring containment hydrogen concentration. The staff sees value in maintaining the capability to obtain grab samples for complementing the information from the hydrogen monitors in the long term (i.e., by confirming the indications from the monitors and providing hydrogen measurements for concentrations outside the range of the monitors). As previously mentioned, the licensees contingency plan (see item 4.1) for obtaining highly radioactive samples will include sampling of the containment atmosphere and may, if deemed necessary and practical by the appropriate decision-makers, be used to supplement the safety-related hydrogen monitors.

4.0 VERIFICATIONS AND COMMITMENTS As requested by the staff in the notice of availability for this CLIIP topic, the licensee has addressed the following plant-specific verifications and commitments.

4.1 Each licensee should verify that it has, and make a regulatory commitment to maintain (or make a regulatory commitment to develop and maintain), contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere.

The licensee has verified that it has contingency plans for obtaining and analyzing highly radioactive samples from the RCS, containment sump, and containment atmosphere. The licensee has committed to maintain the contingency plans within its Chemistry Department implementing procedures. The licensee stated in its application that it has implemented this regulatory commitment.

4.2 Each licensee should verify that it has, and make a regulatory commitment to maintain (or make a regulatory commitment to develop and maintain), a capability for classifying fuel damage events at the Alert level threshold (typically this is 300

.Ci/ml dose equivalent iodine). This capability may utilize the normal sampling system and/or correlations of sampling or letdown line dose rates to coolant concentrations.

The licensee has verified that it has the capability for classifying fuel damage events at the Alert level threshold. The licensee has committed to maintain the capability for the Alert classification within its Emergency Plan implementing procedures. The licensee stated in its application that it has implemented this regulatory commitment.

4.3 Each licensee should verify that it has, and make a regulatory commitment to maintain (or make a regulatory commitment to develop and maintain), the capability to monitor radioactive iodines that have been released to offsite environs.

The licensee has verified that it has the capability to monitor radioactive iodines that have been released to offsite environs. The licensee has committed to maintain the capability for monitoring iodines within its Emergency Plan implementing procedures. The licensee stated in its application that it has implemented this regulatory commitment.

The NRC staff finds that the licensees administrative processes, including its commitment management program, provide reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitments. Should the licensee choose to incorporate a regulatory commitment into the emergency plan, final safety analysis report, or other document with established regulatory controls, the associated regulations would define the appropriate change control and reporting requirements. The staff has determined that the commitments do not warrant the creation of regulatory requirements which would require prior NRC approval of subsequent changes. The NRC staff has agreed that Nuclear Energy Institute 99-04, Revision 0, Guidelines for Managing NRC Commitment Changes, provides reasonable guidance for the control of regulatory commitments made to the NRC staff (see Regulatory Issue Summary 2000-17, Managing Regulatory Commitments Made by Power Reactor Licensees to the NRC Staff, dated September 21, 2000). The commitments should be controlled in accordance with the industry guidance or comparable criteria employed by a specific licensee. The staff may choose to verify the implementation and maintenance of these commitments in a future inspection or audit.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (66 FR 55011). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: W. Reckley Date: January 8, 2002