ML062490489

Technical Specification Amendment Request to Relocate Specifications Not Meeting the Criteria of 10 CFR 50.36(c)(2)(ii), Revision 1,

ML062490489
Person / Time
Site:	Pilgrim
Issue date:	08/30/2006
From:	Balduzzi M Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.06.073, TAC MC5421
Download: ML062490489 (14)

Text

'ýEn teWg Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Michael A. Balduzzi Site Vice President August 30, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Technical Specification Amendment Request to Relocate Specifications Not Meeting the Criteria of 10 CFR 50.36(c)(2)(ii), Revision 1.

(TAC MC5421)

REFERENCE 1. Entergy Letter to NRC 2.04.104, 'Technical Specification Amendment Request to Relocate Various Specifications Not Meeting the Criteria of 10 CFR 50.36(c)(2)(ii), dated December 14, 2004. (TAC MC5421)

LETTER NUMBER: 2.06.073

Dear Sir or Madam:

By this letter, Entergy proposes to revise the license amendment request that was submitted to the NRC by Reference 1. The proposed revision reduces the scope of the original license amendment request by reducing the number of individual items proposed for relocation. There are no.new items proposed for relocation in this revision therefore, the No Significant Hazards Consideration performed for the original license amendment request is unaffected and remains bounding.

Commitments made in this letter are contained in Attachment 3. Entergy will implement the amendment within 90 days following NRC approval.

Ifyou have any questions or require additional information, please contact Bryan Ford at (508) 830-8403.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the L7Y 'of ,7' C 2" 2006.

Sincerely, Michael A. Balduzzi ERS/dm I400 I

Entergy Nuclear Operations, Inc. Letter Number: 2.06.073 Pilgrim Nuclear Power Station Page 2 Attachments: 1. Evaluation of the Proposed Change (4 pages)

2. Mark-up of Technical Specification pages (4 pages)

3. List of Regulatory Commitments (1 page) cc: Mr. James Shea, Project Manager Mr. Robert Walker, Director Plant Licensing Branch I-1 Massachusetts Department of Public Division of Operator Reactor Licensing Health Office of Nuclear Reactor Regulation Radiation Control Program U.S. Nuclear Regulatory Commission 90 Washington Street, 2 nd Floor One White Flint North O-8C2 Dorchester, MA 02121 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 Ms Cristine McCombs, Director U.S. Nuclear Regulator Commission Mass. Emergency Management Agency 475 Allendale Road 400 Worcester Road King of Prussia, PA 19406 Framingham, MA 01702 Senior Resident Inspector Pilgrim Nuclear Power Station

ATTACHMENT 1 Evaluation of the Proposed Chanae

Subject:

Technical Specification Amendment Request to Relocate Specifications Not Meeting the Criteria of 10 CFR 50.36(c)(2)(ii)

1. DESCRIPTION

2. PROPOSED CHANGE

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Environmental Consideration

6. PRECEDENTS

7. REFERENCES

Letter 2.06.073 Page 1 of 4 Technical Specification Amendment Request to Relocate Specifications Not Meeting the Criteria of 10 CFR 50.36(c)(2)(ii)

1. DESCRIPTION Pursuant to 10 CFR 50.90, Entergy proposes to amend the Technical Specifications (TS) for Pilgrim Nuclear Power Station. The proposed change relocates the structural integrity requirements contained in TS 3/4.G to the Pilgrim UFSAR.

2. PROPOSED CHANGE The structural integrity requirements of TS 3.6.G and 4.6.G are proposed for relocation to the UFSAR. The TS and TS Bases pages marked up to reflect the proposed changes are contained in Attachment 2. The marked up TS Bases are provided for information only.

3. BACKGROUND The Pilgrim Nuclear Power Station Inservice Inspection Program conforms to the requirements of 10 CFR 50.55a(g). Where practical, the inspection of ASME Section Xl Class 1, 2, and 3 components conforms to the edition and addenda of Section XI of the ASME Boiler and Pressure Vessel Code required by 10 CFR 50.55a(g). When implementation of an ASME Code required inspection is determined to be impractical for PNPS, a request for relief from the inspection requirement is submitted to the NRC in accordance with 10 CFR 50.55a(g)(5)(iii).

Requests for relief from the ASME Code inspection requirements are submitted to the NRC prior to the beginning of each 10-year inspection interval for which the inspection requirement is known to be impractical. Requests for relief from inspection requirements that are identified to be impractical during the course of the inspection interval are reported to the NRC throughout the inspection interval.

4. TECHNICAL ANALYSIS Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include the TS as part of the license. The Commission's regulatory requirements related to the content for the TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in eight specific categories.

The categories are (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.

However, the regulation does not specify the particular requirements to be included in a plant's TS.

The Commission amended 10 CFR 50.36 (60 FR 36593, July 19,1995), and codified four criteria to be used in determining whether a particular matter is required to be included in a limiting condition for operation (LCO), as follows: (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or

Letter 2.06.073 Page 2 of 4 actuates to mitigate a design-basis accident or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier; or (4) a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety. LCOs and related requirements that fall within or satisfy any of the criteria in the regulation must be retained in the TS, while those requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The PNPS UFSAR is one such licensee-controlled document.

Consistent with these criteria, Entergy proposes to relocate the structural integrity requirements from the PNPS TS to the UFSAR. The four criteria of 10 CFR 50.36 are addressed for the proposed change.

(1) TS 3.6.G and 4.6.G establish the programmatic elements for conducting ASME Code Class 1, 2, and 3 component inspections by reference to Section Xl of the ASME Code.

The safety basis for establishing programmatic requirements on structural integrity in TS relates to prevention of component degradation and continued long-term maintenance of acceptable structural conditions. Therefore, structural integrity of safety systems are not operational limits that are an initial assumption of any DBA or transient analysis.

Additionally, the inspections stipulated by this specification are not used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

(2) The inspections stipulated by TS 3.6.G and 4.6.G do not monitor process variables that are initial assumptions in a DBA or transient analysis.

(3) The ASME Code Class 1, 2, and 3 components inspected per TS 3.6.G and 4.6.G are assumed to function to mitigate accidents. Their capability to perform this function is addressed by other TS. TS 3.6.G and 4.6.G, however, only specifies inspection requirements for these components. Therefore, Criterion 3 is not satisfied.

(4) The TS 3.6.G and 4.6.G requirement is currently covered by 10 CFR 50.55a and the PNPS Inservice Inspection Program. Duplicating regulatory requirements in TS is not a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

The structural integrity requirements of TS 3.6.G and 4.6.G requirements will be relocated to the UFSAR. Any changes to these requirements will be strictly controlled by the provisions of 10 CFR 50.59, as well as 10 CFR 50.55a(g).

Letter 2.06.073 Page 3 of 4

5. REGULATORY SAFETY ANALYSIS 5.1 No Siqnificant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) is proposing to modify the Pilgrim Technical Specifications (TS) to relocate the structural integrity requirements from the TS to the Final Safety Analysis Report (FSAR) or TS Bases. These requirements do not meet the criteria for inclusion in the TS as presented in 10 CFR 50.36(c)(2)(ii).

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed relocation is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation. The associated instrumentation and inspections are not assumed to be an initiator of any analyzed event, nor are these limits assumed in the mitigation of consequences of accidents. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed change to relocate current TS requirements to the FSAR, consistent with regulatory guidance and previously approved changes for other stations, is administrative in nature. The change does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the Technical Specifications to the FSAR. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Letter 2.06.073 Page 4 of 4 Based on the above, Pilgrim concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Environmental Consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii)a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the proposed amendment.

6. PRECENDENTS The NRC has approved similar changes (e.g., relocation of specifications which do not meet the criteria of 10 CFR 50.36(c)(2)(ii)) in a number of amendments. An Example is James A. FitzPatrick conversion to Standard TS, Amendment 274, dated July 3, 2002.

7. REFERENCES

1. Entergy Letter to NRC 2.04.104, 'Technical Specification Amendment Request to Relocate Various Specifications Not Meeting the Criteria of 10 CFR 50.36(c)(2)(ii), dated December 14, 2004. (TAC MC5421)

2. NUREG-1433, Rev. 3, "Standard Technical Specifications, General Electric Plants, BWR/4"

ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATION AND BASES PAGES (4 pages)

TOC page ii TS page 3/4.6-8 TS Bases page B3/4.6-11 TS Bases page B3/4.6-12

TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 3/4.6-1 A. Thermal and Pressurization Limitations A 3/4.6-1 B. Coolant Chemistry B 3/4.6-3 C. Coolant Leakage C 3/4.6-4 D. Safety and Relief Valves D 3/4.6-6 E. Jet Pumps E 3/4.6-7 F. Recirculation Loops Operating e- F 3/4.6-8 G. Str,'turaI Intr*ity Ga O'4.,-8 BASES B3/4.6-1 3.7 CONTAINMENT SYSTEMS 4.7 3/4.7-1 A. Primary Containment A 3/4.7-1 B. Standby Gas Treatment System and Control B Room High Efficiency Air Filtration System 3/4.7-11 C. Secondary Containment C 3/4.7-16 BASES B3/4.7-1 3.8 PLANT SYSTEMS 4.8 3/4.8-1 3.8.1 Main Condenser Offgas 4.8.1 3/4.8-1 3.8.2- MechanicalVacuum Pump-- - .. ....... .- 4.8.2-............... 3/4.8-2--

BASES B3/4.8-1 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 3/4.9-1 A. Auxiliary Electrical Equipment A 3/4.9-1 B. Operation with Inoperable Equipment B 3/4.9-4 BASES B3/4.9-1 3.10 CORE ALTERATIONS 4.10 3/4.10-1 A. Refueling Interlocks A 3/4.10-1 B. Core Monitoring B 3/4.10-1 C. Spent Fuel Pool Water Level C 3/4.10-2 D. Multiple Control Rod Removal D 3/4.10-2 BASES B3/4.10-1 3.11 REACTOR FUEL ASSEMBLY 4.11 3/4.11-1 A. Average Planar Linear Heat Generation Rate A (APLHGR) 3/4.11-1 B. Linear Heat Generation Rate (LHGR) B 3/4.11-2 C. Minimum Critical Power Ratio (MCPR) C 3/4.11-2 D. Power/Flow Relationship During Power D Operation 3/4.11-4 BASES B3/4.11-1 PNPS ii Amendment No.--7, -240-

I 11MITIKII%- eK1r%1T1fK1Q Cfo nOCOAT1 Ki O"MIC11 I A fC OCf%"1O=AACM1M I*-EIIulI ElU~ Ud*I./IuI I DIJIl,1 Fvrll '.V IF~

- I FIUIlI I *JUJI1 V ~IEnLIL-- '.UII rlIU I*I-¥1 I *.J 3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SYSTEM BOUNDARY (Cont)

F. Recirculation Loops Operatina F. Recirculation Loops Operating During operation in the Run and Startup Recirculation pump speeds shall be Modes, at least one recirculation pump shall checked and logged at least once per day.

be operating.

1. Whenever both recirculation pumps are in operation, pump speeds shall be maintained within 10% of each other when power level is greater than 80%

and within 15% of each other when power level is less than or equal to 80%.

2. Whenever a single recirculation loop is operating, the following limits are applied when the associated LCO is applicable:

a) LCO 3.11.A, "Average Planar Linear Heat Generation Rate (APLHGR)," single loop operation limits specified in the COLR, b) LCO 3.11.C, "Minimum Critical

-Power Ratio (MCPR)," single loop operation limits specified in the COLR, and c) LCO 3.1, "Reactor Protection System," Average Power Range Monitor High Flux function, trip level setting for the flow bias function is reset for single loop operation per Table 3.1.1.

3. Ifthe requirements of Specification 3.6.F.1 or 3.6.F.2 are not met, restore

. compliance within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If compliance is not restored or with no l recirculation pumps in operation the reactor shall be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

G. "tctural Inteqrity "-G. St *ural Inteqdrit

1. Th tructural integrity of the primary Inservic inspection of components shall be syste oundary shall be maintained at performed accordance with the PNPS the level uired by the ASME Boiler Inservice Ins ion Program. The results and Pressure essel Code,Section XI obtained from co liance with this program

'Rules for Inserv lnspection of will be evaluated a e completion of each Nuclear Power Plan mponents," ten year interval. The onclusions of this Articles IWA, IWB, IWO, D and IWF evaluation will be review with the NRC.

and mandatory appendic required by 10CFR5O.55a(g), except re specific relief has been granted he NRC pursuant to 10CFR5O.55a(g)(

Amendment No. 19, 93,133,4+9-1 3/4.6-8

3/4. PRIMARY SYSTEM BOUNDARY (Cant The Pilgrim clear Power Station Inservice Inspection Program nforms to the requirements of 10CFR5O.55 ).Where practical, the inspection of ASME Secti Xl Class 1,2, and 3 components confo s to the edition and addenda of Section XI of the ME Boiler and Pressure Vessel Co required by 10CFR5O.55a(g). When implementati of an ASME Code required inspection has een determined to be impractical for PNPS, a requ t for relief from the inspection requiremen Is submitted to the NRC in accordance with 10CFR .55a(g)(5)(iii).

Requests for relief from the ASCode inspection requirements will be submitted t the NRC prior to the beginning of ea ch 10 r inspection interval for which the inspection requi ment is known to be impractical. Requests relief from inspection requirements which are ide 'fied to be impractical during the course of the i ection interval will be reported to the NRC on an annual basis throughout the inspection int al.

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ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS

List of Regulatory.Commitments The following table identifies those actions committed to by Pilgrim in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE Relocate specified requirements to the Within 90 days of license amendment approval.

UFSAR.