ML042950619

Application for Technical Specification Improvement to Extend the Inspection Interval for Reactor Coolant Pump Flywheels Using the Consolidated Line Item Improvement Process

ML042950619
Person / Time
Site:	Harris, Crystal River
Issue date:	10/15/2004
From:	Hinnant C Progress Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PERAS04-102
Download: ML042950619 (15)

Text

j Progress Energy C.S. IHinnant Senior Vice President and Chief Nuclear Officer Progress Energy. Inc.

SERIAL: PERAS04-102 10 CFR 50.90 October 15, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302/LICENSE NO. DPR-72 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT TO EXTEND THE INSPECTION INTERVAL FOR REACTOR COOLANT PUMP FLYWHEELS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Carolina Power & Light Company doing business as Progress Energy Carolinas, Inc. (PEC), and Florida Power Corporation doing business as Progress Energy Florida, Inc. (PEF), are submitting license amendment requests for the Facility Operating Licenses and Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP), and Crystal River Unit 3 Nuclear Generating Plant (CR3). The proposed license amendments will extend the Reactor Coolant Pump (RCP) motor flywheel examination frequency from the currently approved 3-year (HNP only), and 10-year inspection intervals (HNP and CR3), to an interval not to exceed 20 years. The changes are consistent with Industry/Technical Specifications Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-421, "Revision to RCP Flywheel Inspection Program (WCAP-15666)." The availability of this TS improvement was announced in the Federal Register on October 22, 2003, as part of the Consolidated Line Item Improvement Process (CLIIP). HNP is a Westinghouse plant. CR3 is a Babcock and Wilcox plant, not specifically addressed by the NRC Notice of Availability published on October 22, 2003 (68 FR 60422), however, it is bounded by the analyses performed in WCAP-15666.

Attachment 1 provides a description of the proposed changes and the requested confirmation of applicability. Attachment 2 provides the respective Unit's existing TS pages marked-up to show the proposed changes. Attachment 3 provides revised, clean TS pages.

P.O.Box 1551 PEB 1323 Raleigh, NC 27602 T> 919.546.4222 F> 919.546.5473

Document Control Desk PERAS 04-102/ Page 2 These proposed changes have been reviewed and recommended by the Plant Nuclear Safety Committees at each of the respective stations.

PEC and PEF request that the proposed license amendments be reviewed by a single NRC reviewer, and approved by March 31, 2005, with the amendments being implemented within 90 days after approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina and Florida State Officials.

Please refer any questions regarding this submittal to Mr. Tony Groblewski, Supervisor - Corporate Regulatory Affairs, at (919) 546-4579.

Sincerely, C. S. Hinnant CMI Attachments: 1. Description and Assessment

2. Proposed Technical Specification Changes

3. Revised Technical Specification Pages C. S. Hinnant, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Progress Energy Carolinas, Inc. (PEC) and Progress Energy Florida, Inc. (PEE).

N (Seal)

My commission expires: SJ 59 0 0 V

Document Control Desk PERAS 04-102/ Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission AT'IN: Mr. Chandu P. Patel (Mail Stop OWFN 8H12) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 USNRC Senior Resident Inspector Shearon Harris Nuclear Power Plant, Unit No. 1 USNRC Senior Resident Inspector Crystal River Unit 3 Nuclear Generating Plant Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Ms. Beverly 0. Hall, Section Chief Radiation Protection Section, Division of Environmental Health North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221 Mr. William A. Passetti, Chief Department of Health, Bureau of Radiation Control 2020 Capital Circle, SE Bin #C21 Tallahassee, FL 32399-1741 Chairman Board of County Commissioners Citrus County 110 N. Apopka Avenue, Second Floor Inverness, FL. 34450-4245

ATTACHMENT 1 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT TO EXTEND THE INSPECTION INTERVAL FOR REACTOR COOLANT PUMP FLYWHEELS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS DESCRIPTION AND ASSESSMENT

1.0 INTRODUCTION

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Carolina Power & Light Company doing business as Progress Energy Carolinas, Inc.

(PEC), and Florida Power Corporation doing business as Progress Energy Florida, Inc. (PEF),

propose changes to the Technical Specifications (TS), for the following Operating Licenses:

Progress Energy Carolinas, Inc. (PEC)

SHEARON HARRIS NUCLEAR POWER DOCKET NO. 50400 PLANT, UNIT 1 LICENSE NO. NPF-63 Progress Energy Florida, Inc. (PEF)

CRYSTAL RIVER UNIT 3 NUCLEAR DOCKET NO. 50-302 GENERATING PLANT LICENSE NO. DPR-72 The proposed License amendments would modify the TS requirements for the Reactor Coolant Pump (RCP) Motor Flywheel Inspection Program to extend the examination frequency for the RCP motor flywheels from the currently approved 3-year (Shearon Harris Nuclear Power Plant, Unit 1 only) and 10-year inspection intervals (Shearon Harris Nuclear Power Plant, Unit 1 and Crystal River Unit 3 Nuclear Generating Plant), to an interval not to exceed 20 years. The changes are consistent with Industry/Technical Specification Task Force (TSTF) Standard Technical Specifications Change Traveler, TSTF-421, "Revision to RCP Flywheel Inspection Program (WCAP-15666)." The availability of this TS improvement was announced in the Federal Register on October 22, 2003, as part of the Consolidated Line Item Improvement Process (CLIIP).

2.0 DESCRIPTION

OF PROPOSED AMENDMENT Consistent with the NRC-approved TSTF-421, the proposed TS changes include the following revisions to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), and the Crystal River Unit 3 Nuclear Generating Plant (CR3),TS.

2.1 HNP The proposed changes affect HNPTS Surveillance Requirement 4.4.10, which will be revised to read:

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"Each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PIT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals."

The proposed HNP TS changes are indicated in Attachments 2 and 3.

2.2 CR3 The proposed changes affect CR3 TS Section 5.6.2.8.c, "Inservice Inspection Program," which will be revised to read:

"Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every twenty years."

The proposed CR3 TS changes are indicated in Attachments 2 and 3.

3.0 BACKGROUND

The background for this application is adequately addressed by the NRC Notice of Availability published on October 22, 2003 (68 FR 60422), NRC Notice for Comment published on June 24, 2003 (68 FR 37590), TSTF-421, WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination," and the related NRC safety evaluation (SE) dated May 5, 2003.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE 4.1 HNP The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on October 22, 2003 (68 FR 60422), NRC Notice for Comment published on June 24, 2003 (68 FR 37590), TSTF-421, WCAP-15666, and the related NRC SE.

4.2 CR3 CR3 is a Babcock and Wilcox plant, not specifically addressed by the NRC Notice of Availability, however, it is bounded by the analyses performed in WCAP-15666. The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on October 22,2003 (68 FR 60422), NRC Notice for Comment published on June 24, 2003 (68 FR 37590), TSTF-421, WCAP-15666, and the related NRC SE.

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5.0 TECHNICAL ANALYSIS

5.1 HNP Progress Energy Carolinas, Inc. (PEC), has reviewed the model SE published on June 24, 2003 (68 FR 37590), and verified its applicability as part of the CLIIP. This verification included a review of the NRC staff's model SE, as well as the information provided to support TSTF-421 (including WCAP-15666 and the related SE dated May 5, 2003). PEC has concluded that the justifications presented in the TSTF proposal and the model SE prepared by the NRC staff are applicable to HNP and justify this amendment for the incorporation of the changes to the HNP TS.

5.2 CR3 CR3 is a Babcock and Wilcox (B&W) plant, not specifically addressed by the NRC Notice of Availability published on October 22, 2003 (68 FR 60422), however, it is bounded by the analyses performed in WCAP-115666 (Reference 6). Progress Energy Florida, Inc. (PEF), has reviewed the model SE published on June 24, 2003 (68 FR 37590), and verified its applicability as part of the CLIIP. This verification included a review of the NRC staff's model SE, as well as the information provided to support TSTF-421 (including WCAP-15666 and the related SE dated May 5, 2003). PEF has concluded that the justifications presented in the TSTF proposal and the model SE prepared by the NRC staff are applicable to CR3 and justify this amendment for the incorporation of the changes to the CR3 Improved Technical Specifications (ITS). This conclusion is supported by the following statements.

The NRC Safety Evaluation (SE) issued for CR3 on August 31, 1998 (Reference 7), allowed the extension of the CR3 RCP motor flywheel inspection interval based on the Westinghouse topical report WCAP-14535A, 'Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination." This report included all domestic operating Westinghouse plants and several B&W plants, including CR3 which is one of the Group 16 plants analyzed in WCAP-14535A. The NRC SE confirmed that the four RCP motor flywheels at CR3 made of A516 material were included in WCAP-14535A, Group 16, and concluded that the A516 material has adequate fracture toughness for this application. The fifth RCP motor flywheel, at CR3 is made of A508 material. Based on material properties, the NRC concluded that A508 is as tough as the SA 533B material used in the Westinghouse RCP motor flywheels. Therefore, the NRC staff determined that the analysis and conclusions in the WCAP-14535A for SA 533B material are applicable to the CR3 RCP motor flywheels, and approved the inspection interval extension (i.e., perform either volumetric or surface examination of the RCP motor flywheels once every ten years).

WCAP-1 5666, "Extension of Reactor Coolant Pump Motor Flywheel Examination" (Reference 4), justifies the extension of the RCP motor flywheel inspection interval for all domestic Westinghouse plants from 10 years to 20 years, based on the WCAP-14535A and additional risk evaluations. WCAP-15666 has been reviewed and was approved by the NRC on May 5,2003 (Reference 5). CR3 is included in WCAP-15666, Table 2-3 (Flywheel Inspection Results from MUHP-5042 Study). The CR3 RCP motor flywheels design conditions have been evaluated for applicability to WCAP-15666, Table 2-2 (Summary of Westinghouse Domestic Plant RCP A1-3

Motor Flywheel Information), and Table 3-8 (Cumulative Probability of Failure over 40 and 60 Years with and without Inservice Inspection). The CR3 RCP miotor flywheels are similar to the analyzed Westinghouse RCP motor flywheels by comparison of the flywheels diameter, bore, pump & motor inertia, design limiting speed, and previous material acceptability. The CR3 RCP motor flywheels design basis maximum speed is 1800 rpm. This is bounded by the value of 3321 rpm, which is the highest analyzed speed for RCP motor flywheels at Westinghouse domestic plants, listed in WCAP-15666, Table 3-8.

The WCAP-15666 risk evaluations assumptions, which had to be verified by B&W plants in order for Westinghouse to confirm that the WCAP applies to them, are identified below, together with CR3 plant specific data.

1 - Probability of core damage given a RCP motor flywheel failure is 1.0 in WCAP-15666. This is conservative and bounding for CR3.

2 - Initiating event frequency for a plant trip is estimated as one event per year based on a generic industry trip rate, in WCAP-15666. The current PRA model initiating event frequency for CR3 is slightly higher (i.e., 1.5 per year) than this number for the total frequency of all plant trip initiators. The risk relationship to the initiating event frequency is linear and since the calculated risk in WCAP-15666 is 1.2E-8, doubling the trip frequency would make the risk 2.4E-8, which is significantly below the limit of 1E-7 (RG 1.174 guidance).

3 - The probability of Loss of Offsite Power (LOOP), or loss of power to the RCP following a plant trip or Loss of Coolant Accident (LOCA) is 1.4E-2 in WCAP-15666. CR3 has not experienced a loss of offsite power due to a reactor trip. CR3 design provides two separate switchyards for the main generator connection and the power supply to the safety busses. This number is therefore conservatively bounding.

4 - The mean value for the frequency of large LOCA evaluated in WCAP-15666 is 2E-06/yr, which is comparable to that of a 5E-06/yr given for a "Large Pipe Break LOCA" in NUREG/CR-5750 (Reference 8). The large LOCA frequency for CR3 is 5E-06/year, and it is based on NUREG/CR-5750.

5 - RCP flywheel failure probabilities are shown in Table 3-8 of WCAP-15666. The RCP motor flywheels at CR3 have inertia of 70,000 lb-ft 2 and a maximum speed of 1800 rpm in the event of a large LOCA with no electric power. This event is bounded by the data for RCP motor flywheels described in WCAP-15666 Table 2-2, with inertias of 70,000 - 110,000 lb-ft2 , and an analyzed maximum speed of 3321 rpm. Material properties of the CR3 RCP motor flywheels have been previously evaluated by the NRC and found acceptable (Reference 7). The current CR3 spare RCP motor flywheel has been inspected by Ultrasonic Testing (UT), and Magnetic Particle Testing (MT), during refurbishment maintenance in August 1997. All four CR3 RCP motor flywheels in service have been recently inspected by UT. Two were inspected during the refueling outage of October 2001, and the other two were inspected during the refueling outage of October 2003. There were no indications of material degradation detected during these inspections.

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6 - Failure scenarios are described in WCAP-15666, Section 3.2. The RCP motor flywheel failure scenarios outlined in Section 3.2 of WCAP-15666 (i.e.,failiure causing plant trip, failure after reactor trip with electric powver, failure after trip without electric power, large LOCA without electric power) are applicable to CR3 RCP motor flywheels, with no additional failure scenarios.

Plant specific evaluation of the CR3 RCP motor flywheels confirmed that WCAP-15666 is applicable to CR3. The applicability of WCAP-15666 to CR3 is documented in the WOG letter to the NRC dated September 8, 2003 (Reference 6). Therefore, the change in risk from extending the inspection interval for the CR3 RCP motor flywheels from 10 years to 20 years is significantly below the acceptance criteria of RG 1.174.

6.0 REGULATORY ANALYSIS

A description of the proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC notices related to the CLIIP, TSTF-421, topical report WCAP-15666, and the associated SE.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION Progress Energy Carolinas, Inc. (PEC), and Progress Energy Florida, Inc. (PEF), have reviewed the proposed no significant hazards consideration determination published on June 24, 2003 (68 FR 37590) as part of the CLIP. PEC and PEF have concluded that the proposed determination presented in the notice is applicable to HNP and CR3 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

8.0 ENVIRONMENTAL EVALUATION Progress Energy Carolinas, Inc. (PEC), and Progress Energy Florida, Inc. (PEF), have reviewed the environmental evaluation included in the model SE published on June 24, 2003 (68 FR 37590), as part of the CLIIP. PEC and PEF have concluded that the staff's findings presented in that evaluation are applicable to HNP and CR3. The evaluation is hereby incorporated by reference for this application.

9.0 PRECEDENT This application is being made in accordance with the CL1IP. PEC and PEF are not proposing variations or deviations from the TS changes described in TSTF-421 or the NRC staff's model SE published on June 24, 2003 (68 FR 37590).

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10.0 REFERENCES

1. Federal Register Notice: Notice of Availability of Model Application Concerning Technical Specification Improvement Regarding Extension of Reactor Coolant Pump Motor Flywheel Examination for Westinghouse Plants Using the Consolidated Line Item Improvement Process, published October 22, 2003 (68 FR 60422).

2. Federal Register Notice: Notice of Opportunity to Comment on Model Safety Evaluation on Technical Specification Improvement Regarding Extension of Reactor Coolant Pump Motor Flywheel Examination for Westinghouse Plants Using the Consolidated Line Item Improvement Process, published June 24, 2003 (68 FR 37590).

3. Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF421, "Revision to RCP Flywheel Inspection Program (WCAP-15666)," Revision 0, November 2001.

4. WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination," July 2001.

5. NRC letter dated May 5, 2003, from H. Berkow to R. Bryan (WOG) transmitting Safety Evaluation of WCAP-15666.

6. WOG letter WOG-03-467, dated September 8, 2003, from R. H. Bryan (WOG) to W. D. Reckley (NRC), Applicability of WCAP-15666-NP, Rev. 0, "Extension of Reactor Coolant Pump Motor Flywheel Examination," to B&W NSSS Plants.

7. NRC letter dated August 31, 1998, from L. A. Wiens to J. P. Cowan (Florida Power Corporation), "Crystal River Unit 3- Staff Evaluation and Issuance of Amendment re:

Reactor Coolant Pump Flywheel Inspection Requirements" (TAC NO. MA2145).

8. WCAP-15666-A, Rev. 1, (Non-Proprietary) "Extension of Reactor Coolant Pump Motor Flywheel Examination," October 2003.

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ATTACHMENT 2 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT TO EXTEND THE INSPECTION INTERVAL FOR REACTOR COOLANT PUMP FLYWHEELS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS PROPOSED TECHNICAL SPECIFICATION CHANGES Progress Energy Carolinas, Inc. (PEC)

SHEARON HARRIS NUCLEAR POWER DOCKET NO. 50-400 PLANT, UNIT 1 LICENSE NO. NPF-63 Progress Energy Florida, Inc. (PEF)

CRYSTAL RIVER UNIT 3 NUCLEAR l DOCKET NO. 50-302 GENERATING PLANT LICENSE NO. DPR-72

REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1. 2. and 3 components shall be maintained in accordance with Specification 4.4.10.

APPLICABILITY: All MODES.

ACTION:

a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements. restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 500 F above the minimum temperature required by NDT considerations.

b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200 0F.

c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements. restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.

SURVEILLANCE REQUIREMENTS 4.4.10 kinn to the requirements of Specification 4.0.5 s ch reactor coolant pump flywheel shall be inspected per recommdions a io of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1. August 1975.

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SHEARON HARRIS - UNIT 1 3/4 4-43 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals I-I 5.6.2.7 Not Used 5.6.2.8 Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, 3, MC and CC components, including applicable supports. The program shall include the following:

a. Provisions that inservice inspection of ASME Code Class 1, 2, 3, MC and CC components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

b. The provisions of SR 3.0.2 are applicable to the frequencies for performing inservice inspection activities;

c. Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every &wenty ten years. The inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination for exposed surfaces of the disassembled flywheels. The recommendations delineated in Regulatory Guide 1.14, Positions 3, 4, and 5 of Section C.4.b shall apply.

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

(conti nued)

Crystal River Unit 3 5 .0-1 Amendment No. Stir

ATTACHMENT 3 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT TO EXTEND THE INSPECTION INTERVAL FOR REACTOR COOLANT PUMP FLYWHEELS USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS REVISED TECHNICAL SPECIFICATION PAGES Progress Energy Carolinas, Inc. (PEC)

SHEARON HARRIS NUCLEAR POWER DOCKET NO. 50400 PLANT, UNIT 1 LICENSE NO. NPF-63 Progress Energy Florida, Inc. (PEF)

CRYSTAL RIVER UNIT 3 NUCLEAR DOCKET NO. 50-302 GENERATING PLANT LICENSE NO. DPR-72

REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1. 2. and 3 components shall be maintained in accordance with Specification 4.4.10.

APPLICABILITY: All MODES.

ACTION:

a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations.

b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200 0F.

c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.

SURVEILLANCE REQUIREMENTS 4.4.10 Each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14.

Revision 1. August 1975. In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.

SHEARON HARRIS - UNIT 1 3/4 4-43 Amendment No.

.,. 6 I . I .' , -

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.7 Not Used 5.6.2.8 Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, 3, MC and CC components, including applicable supports. The program shall include the following:

a. Provisions that inservice inspection of ASME Code Class 1, 2, 3, MC and CC components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

b. The provisions of SR 3.0.2 are applicable to the frequencies for performing inservice inspection activities;

c. Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every twenty years. The I inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination for exposed surfaces of the disassembled flywheels. The recommendations delineated in Regulatory Guide 1.14, Positions 3, 4, and 5 of Section C.4.b shall apply.

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

(continued)

Crystal River Unit 3 5 .0-1 Amendment No.