L-08-325, License Amendment Request No. 08-18 to Incorporate Technical Specification Task Force Travelers 479 and 497

License Amendment Request No. 08-18 to Incorporate Technical Specification Task Force Travelers 479 and 497

ML083390728
Person / Time
Site:	Perry
Issue date:	11/18/2008
From:	Bezilla M FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-325
Download: ML083390728 (24)

Text

FENOC %10 Perry Nuclear Power Ston CenterRoad FirstEnergyNuclear OperatingCompany Perry,Ohio 44081 Mark B. Bezilla 440-280-5382 Vice President Fax: 440-280-8029 November 18, 2008 L-08-325 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 License Amendment Request No. 08-18 to Incorporate Technical Specification Task Force Travelers 479 and 497 Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) hereby requests an amendment to the operating license for the Perry Nuclear Power Plant (PNPP). The proposed amendment would modify Technical Specification (TS) 5.5.6 to incorporate Technical Specification Task Force (TSTF) Travelers TSTF-479, "Changes to Reflect Revision of 10 CFR 50.55a," and TSTF 497, "Limit Inservice Testing Program SR

[Surveillance Requirement] 3.0.2 Application to Frequencies of 2 Years; or Less."

The PNPP second 10-year Inservice Testing (IST) interval concludes on May 17, 2009.

Therefore, FENOC requests approval of the proposed amendment by May 17, 2009, to coincide with the beginning of the third 10-year IST interval. There are no regulatory commitments contained in this letter. Ifthere are any questions or additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November .L., 2008.

Sincerely, Mark B. Bezilla

Enclosure:

Evaluation of the Proposed License Amendment Request cc: NRC Region III Administrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board 0C-/7

Evaluation of Proposed License Amendment Page 1 of 7

Subject:

License Amendment Request to incorporate Technical Specification Task Force Traveler (TSTF) 479 and TSTF 497.

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES Attachments:

1. Proposed Technical Specification Changes (Mark Up)

2. Proposed Technical Specification Changes (Re-typed -- For Information)

3. Proposed Technical Specification Bases Pages (Mark Up - For Information)

Perry Nuclear Power Plant License Amendment Request 08-018 Page 2 of 7 1.0

SUMMARY

DESCRIPTION The FirstEnergy Nuclear Operating Company requests Nuclear Regulatory Commission (NRC) review and approval of an amendment to the Perry Nuclear Power Plant (PNPP) Technical Specifications (TS). The proposed amendment updates references to the source of requirements for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves and addresses the applicability of Surveillance Requirement (SR) 3.0.2 to some non-standard pump and valve testing frequencies.

The proposed changes are administrative and are consistent with NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-479, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a," and TSTF 497, Revision 0, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less."

2.0 DETAILED DESCRIPTION The proposed change adopts TSTF-479, Revision 0 to revise references to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code in Technical Specification (TS) 5.5.6, "Inservice Testing Program." The proposed change deletes reference to Section XI of the Code and incorporates reference to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). The proposed change also adopts TSTF-497, Revision 0 to limit Surveillance Requirement (SR) 3.0.2 applicability to normal and accelerated Inservice Testing (IST) Program frequencies specified as two years or less that are not specifically listed in the testing frequencies identified in TS 5.5.6.

2.1 Proposed Changes TS 5.5.6, "Inservice Testing Program," is revised to indicate that the Inservice Testing Program (IST) shall have testing Frequencies applicable to the ASME OM Code.

TS 5.5.6 is also revised to indicate that the provisions of SR 3.0.2 are applicable to some nonstandard frequencies utilized in the IST Program. Specifically, TS 5.5.6.b is revised to state:

"The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities."

Perry Nuclear Power Plant License Amendment Request 08-018 Page 3 of 7 Marked up pages of the affected Technical Specifications are provided in Attachment 1. Re-typed clean Technical Specification pages are provided for information in Attachment 2.

Various sections of the TS Bases will be revised for consistency with the requirements of 10 CFR 50.55a(f)(4). The changes to the affected TS Bases pages will be incorporated in accordance with TS 5.5.11, "Technical Specifications (TS)

Bases Control Program." Marked up pages of the affected Technical Specifications Bases are provided for information in Attachment 3.

2.2 Background

In 1990, the ASME published the initial edition of the ASME OM Code that provides rules for inservice testing of pumps and valves. The ASME OM Code was developed and is maintained by the ASME Committee on Operation and Maintenance of Nuclear Power Plants. The ASME OM Code was developed in response to the ASME Board on Nuclear Codes and Standards directive that transferred responsibility for development and maintenance of requirements for the inservice testing of pumps and valves from the ASME,Section XI, Subcommittee on Nuclear Inservice Inspection to the ASME OM Committee. Therefore, the ASME OM Code is to replace Section XI of the Boiler and Pressure Vessel Code for the inservice testing of pumps and valves. The 1995 edition of the ASME OM Code was incorporated by reference into 10 CFR 50.55a(b).

Since 10 CFR 50.55a(f)(4)(ii) requires that inservice testing during successive 10-year intervals complies with the requirements of the latest edition and addenda of the Code incorporated into 10 CFR 50.55a(b), TS 5.5.6 must be revised to reference the ASME OM Code.

3.0 TECHNICAL EVALUATION

The purpose of the Inservice Testing (IST) program is to assess the operational readiness of pumps and valves, to detect degradation that might affect component OPERABILITY, and to maintain safety margins with provisions for increased surveillance and corrective action. 10 CFR 50.55a defines the requirements for applying industry codes to each licensed nuclear power facility.Section XI of the ASME Code has been revised on a continuing basis over the years to provide updated requirements for the inservice inspection and inservice testing of components. As discussed above, until 1990, the ASME Code requirements addressing the inservice testing of pumps and valves were contained in Section Xl, Subsections IWP (pumps) and IWV (valves). Since the establishment of the 1990 Edition of the OM Code; the rules for inservice testing have been removed from Section XI.

Perry Nuclear Power Plant License Amendment Request 08-018 Page 4 of 7 The Code of record for the PNPP second 10-year IST testing interval, which concludes on May 17, 2009, is the 1989 Edition of ASME Code,Section XI. The PNPP third 10-year IST interval will begin on May 18, 2009, and the IST program for the interval will be in accordance with the 2001 Edition through the 2003 Addenda of the ASME OM Code, as required by 10 CFR 50.55a(f)(4)(ii).

By final rule issued on September 22, 1999, the NRC amended 10 CFR 50.55a(f)(4)(ii) to require licensees to update their IST Program to the latest approved edition of the ASME OM Code. The PNPP TS, 5.5.6, currently references the ASME Boiler and Pressure Vessel Code,Section XI, as the source of the IST Program requirements for ASME Code 1, 2, and 3 components.

Therefore, the proposed changes to TS 5.5.6 are necessary to reference the OM Code versusSection XI.

TS 5.5.6 is also revised to limit the scope of applicability of the provisions of SR 3.0.2 for those IST activities with frequencies that are not specifically listed in the testing frequencies identified in TS 5.5.6. The IST program may have frequencies for testing that are based on risk or other factors and do not conform to the standard testing frequencies. Application of SR 3.0.2 to these other IST frequencies specified as two years or less is consistent with the guidance in NUREG-1482, Revision 1, "Guidelines for Inservice Testing at Nuclear Power Plants," paragraph 3.1.3.

4.0 REGULATORY EVALUATION

The proposed changes are administrative in nature and revise the requirements in TS 5.5.6, "Inservice Testing Program," to update references to the ASME Boiler and Pressure Vessel Code, Section Xl, as the source of requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes delete reference to Section XI of the Code, incorporate reference to the ASME OM Code, and address the applicability of Surveillance Requirement (SR) 3.0.2 to normal and accelerated Inservice Testing Program frequencies specified as two years or less that are not specifically listed in the testing frequencies identified in TS 5.5.6.

4.1 Significant Hazards Consideration The FirstEnergy Nuclear Operating Company (FENOC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment to the PNPP TS by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

Perry Nuclear Power Plant License Amendment Request 08-018 Page 5 of 7 Response: No.

The proposed amendment revises TS 5.5.6, "Inservice Testing Program,"

for consistency with 10 CFR 50.55a(f)(4) requirements regarding inservice testing of pumps and valves. The proposed amendment incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed changes do not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, the proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes do not involve a modification to the physical configuration of the plant. There is no new equipment to be installed or a change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual cumulative occupational exposure. Therefore, the proposed change does not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No.

The proposed amendment revises TS 5.5.6, "lnservice Testing Program,"

for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves. The proposed amendment incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Perry Nuclear Power Plant License Amendment Request 08-018 Page 6 of 7 Based on the above, FENOC concludes that the proposed amendment does not involve a 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.

4.2 Applicable Regulatory Requirements/Criteria NRC regulation, 10 CFR 50.55a, defines the requirements for applying industry codes to each licensed nuclear powered facility. The regulations require that IST of pumps and valves conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the Code incorporated into paragraph (b) of 10 CFR 50.55a 12 months before the start of the 120-month interval.

The proposed amendment ensures FENOC will continue to comply with the requirements of 10 CFR 50.55a.

4.3 Precedent The NRC accepted TSTF-479 in December 2005 and TSTF-497 in October 2006.

Amendments adopting TSTF-479 and TSTF-497 have been approved for the Brunswick Steam Electric Plant (Amendments 247 and 275 - TAC Nos.

MD6916 and MD6917), the Diablo Canyon Power Plant (Amendments 196 and 197 - TAC Nos. MD3975 and MD3976), and the Columbia Generating Station (Amendment 205 - TAC No. MD6209).

4.4 Conclusions In conclusion, based on the considerations discussed above, (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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed change would change a requirement with respect to installation or use of a facility component located within the restricted areas, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change 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

Perry Nuclear Power Plant License Amendment Request 08-018 Page 7 of 7 in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 be prepared in connection with the proposed change.

6.0 REFERENCES

1. U. S. Nuclear Regulatory Commission, "Guidance for Inservice Testing at Nuclear Power Plants," NUREG-1482, Revision 1, January 2005

2. Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-479, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a"

3. Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-497, Revision 0, "Limit Inservice Tresting Program SR 3.0.2 Application to Frequencies of 2 Years or Less"

4. Federal Register Notice: Industry Codes and Standards; Amended Requirements, published September 22, 1999 (64 FR 51370)

5. American Society of Mechanical Engineers Operation and Maintenance Code for Operation and Maintenance of Nuclear Power Plants, :2001 Edition including the OMa-2002 and OMb-2003 Addenda

Attachment 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (Mark Ups)

Page 1 of 3

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program-shall include the__

following:

AM-endaterminolob' for inservice testinQ activities

....... ii Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually. At least once per 366 days Biennially or every 2 years At least once per 731 days

c. The provisions of SR 3.0.3 are applicable to inservice testing activities: and

d. Nothing in the ASME Boiler and Pressure ___Conde shall bE construed to supersede the requlrements of any TS.

5.5.7 Ventilation Filter Testing ProQram (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2.

(continued)

PERRY - UNIT 1 5.0-10 Amendment No.4

INSERT 1 The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities;

Attachment 2 PROPOSED TECHNICAL SPECIFICATION CHANGES (Re-Typed - For Information Only)

Page 1 of 2

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2. and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code)and applicable Addenda as follows:

ASMEOM Code and applicable Required frequencies Addenda terminology for for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Year]y or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

5.5.7 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2.

(continued)

PERRY - UNIT 1 5.0-10 Amendment No.

Attachment 3 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (Mark Ups - For Information Only)

Page 1 of 11

S/RVs B 3.4.4 BASES SURVEILLANCE SR 3.4.4.3 (continued)

REQUIREMENTS The successful performance of the S/RVs tested provides

• reasonable assurance that the remaining installed S/RVs will perform in a similar fashion. After the. S/RVs are replaced, the power-operated actuator of all 19 S/RVs will be uncoupled from the S/RV stem, and cycled to ensure proper operation of the control circuit and actuator. Following cycling, the power-operated actuator is recoupled and the proper positioning of the stem nut is independently verified. This verifies that each S/RV will properly perform its intended function. If the valve actuator fails to operate due only to the failure of the solenoid but iscapable of opening the valve on overpressure, the safety function of theý S/RV is considered OPERABLE.

When removing and replacing the SIRVs, Foreign Material Exclusion controls will be in place to minimize the potential for unwanted materials from entering into any S/RV opening or the piping discharge lines.

SR 3.4.4.2 and the LOGIC SYSTEM FUNCTIONAL TEST performed in SR 3.3.6.4.4 overlap this surveillance to provide complete testing of the assumed safety function The 24 months on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 24 month Frequency wa develoed basedon th2eS/RVtests 9 rggure the ME( .... 'Pl=]Cod k)enn. The-24month Frequency is based on operaTing experien e. and is consistent with a typical industry refueling 1c S@

REFERENCES 1. 4AM oiler and Pressure Vessel Code. SectioneIII.

2. USAR, Chapter 15, Appendix 15B.

3. USAR, Section 15.

4. NRC Safety Evaluation to NEDC-31753P. March 8. 1993.

5P. -[UNT 1 5.42ppRendnX IN PERRY -UNIT 1 B 3.4-22 Revision No. 5

INSERT 1 ASME Code for Operation and Maintenance of Nuclear Power Plants

RCS PIV Leakage B 3.4.6 BASES SURVEILLANCE SR 3.4.6.1 (continued)

REQUIREMENTS The Frequency require rvice Testing Program is within the ASME Code. ,T,Frequency requirement.

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. 10 CFR 50. Appendix A, GDC 55.

4. 424, Rol!e@Pnd PPes,-re Vessel 9ed- Szectins* ]

5. NUREG-0677. "The Probability of Intersystem LOCA:

Impact Due to Leak Testing and Operational Changes,"

May 1980.

6. PNPP - Unit 1, Inservice Test Program.

PERRY - UNIT 1 B 03.4-32 Revision No. 1

"- -ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.7 (continued)

REQUIREMENTS SR 3.5.1.6 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1.6 overlap this Surveillance to provide complete testing of the safety function. The Frequency of 24 months on a STAGGERED TEST BASIS Frequency ensures that both solenoids for each ADS valve power-operated actuator are alternately tested. The Frequency of the required-power-opera ted__*ctuat~nG++/-est i,5 based on the tests required by ASM- Ns implemented by the Inservice c-T~i estig -- ogram6-tSpeciication 5.5.6. The testing Coc* Frequency required by the.Inservice Testing Program is based on operating experience and valve performance. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.5.1.8 This SR ensures that the ECCS RESPONSE TIMES are within limits for each of the ECCS injection and spray. subsystems.

This SR is modified by a note which identifies that the associated ECCS actuation instrumentation is not required to be response time tested. Response time testing of the remaining subsystem components is required. This is supported by Reference 15. Response time testing acceptance criteria are included in Reference 16.

ECCS RESPONSE TIME tests are conducted every 24 months. The 24 month Frequency is based on the need to perform this (continued)

PERRY - UNIT 1 B 3.5-13a Revision No. 5

ECCS-Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS SR 3.5.1.8 (continued)

Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on operating eýperience, and is consistent with a typical industry refueling cycle.

REFERENCES 1. USAR, Section 6.3:2.2.3.

2. USAR. Section 6.3.2.2.4.

3. USAR, Section 6.3.2.2.1.

4. USAR, Section 6.3.2.2.2.

5. USAR, Section 15.6.6.

6. USAR. Section 15.6.4.

7. USAR, Section 15.6.5.

8. 10 CFR 50, Appendix K.

9. USAR, Section 6.3.3.

10. 10 CFR 50.46.

11. USAR, Section 6.3.3.3.

12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), "Recommended Interim Revisions to LCO's for ECCS Components." December 1. 1975.

13. USAR, Section 5.2.2.4.1. EF - .

14. o SG.

15. NEDO-32291, "System Analyses for Elimination of Selected Response Time Testing Requirements,"

January 1994.

16. USAR, Section 6.3, Table 6.3-1.

PERRY - UNIT 1 B 3.5-14 Revision No. 5

LLS Valves B 3.6.1.6 BASES (continued)

SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS Method 2:

The required population of LLS S/RVs tested will be stroke in the relief mode during testing at a qualified offsite facility to verify proper operation of the S/RV.

The successful performance of the S/RVs tested provides reasonable assurance that the remaining installed S/RVs will perform in a similar fashion. After the S/RVs are replaced, the power-operated actuator of all 19 S/RVs will be uncoupled from the S/RV stem, and cycled to ensure proper operation of the control circuit and actuator. Following cycling, the power-operated actuator is recoupled and the proper positioning of the stem nut is independently verified. This verifies that each S/RV will properly perform its intended function. If the valve actuator fails to operate due only to the failure of the solenoid but is capable of opening the valve on overpressure. the safety mode of the S/RV is considered OPERABLE.

When removing and replacing the S/RVs, Foreign Material Exclusion controls will be in place to minimize the potential for unwanted materials from entering into any S/RV opening or the piping discharge lines.

The STAGGERED TEST BASIS Frequency ensures that both solenoids for each LLS valve power-operated actuator are alternately tested. *The 24 Month Frequency of the required power-operated ac m-lesting i based on the tests required by ASRef. s implemented by the 1iie g.Prograaof Specification 5.5.6. The

- testing Frequency required by th Inservice Testing Program is based on operating-experience and valve performance.

Therefore, the Frequency was con luded to be acceptable from a reliability standpoint.

(continued)

PERRY - UNIT I B 3.6-41a Revision No. 5

LLS Valves B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.2 REQUIREMENT

.(continued) The LLS function S/RVs are required to actuate automatically upon receipt of specific initiation signals. A functional test is performed to verify that the mechanical portions (i.e.. solenoids) of the automatic LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.4.4 overlaps this SR to provide complete testing of the-safety function.

The 24 month Frequency is based on the need to perform this Surveillance during a plant outage and the potential for an unplanned transient if the Survei-llance were performed with the reactor at power. The 24 month Frequency is based on operating experience, and is consistent with a typical industry refueling cycle.

This SR is modified by a Note that excludes valve actuation.

This prevents a reactor pressure vessel pressure blowdown.

REFERENCES 1. GESSAR-II. Appendix 3B. Attachment A, Section 3BA.8.

2. USAR, Section 7.6.1.11.

3. Af ltr--. " Xi _ .7_

PERRY - UNIT 1 B 3.6-42 Revision No. 5

RHR Containment .Spray System B 3.6.1.7 BASES SURVEILLANCE SR 3.6.1.7.1 (continued)

REQUIREMENTS A Note has been added to this SR that allows RHR containment spray subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam pressure less than the RHR cut in ermissive pressure in MODE 3, if capable of being manually realigned (remote or local) and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3 if necessary.

SR 3.6.1.7.2 Verifying each RHR pump develops a flow rate 2 5250 gpm with flow through the associated heat exchangers ensures that pump performance has not degraded below the required flow rate during the cycle. It is tested in the suppression pool cooling mode to demonstrate pump OPEFABILITY without spraying down equipment in primary containment. Flow is a normal test of tenfugal pump performance required by the ASME 2). This test confirms one point on the-pump-'sign curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance. and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.

SR 3.6.1.7.3 This SR verifies that each RHR containment spray subsystem automatic valve actuates to its correct position upon receipt of an actual or simulated automatic initiation signal. Actual spray initiation is not required to meet this SR. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.5 overlaps this SR to provide complete testing of the safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is based on (continued)

PERRY - UNIT 1 B 3.6-46 Revision No. 3

RHR Containment Spray System B 3.6.1.7 QBA SES SURVEILLANCE SR 3.6.1.7.3 (continued)

REQUIREMENTS operating experience, and is consistent with a typical industry refueling cycle.

SR 3.6.1.7.4 This Surveillance is performed following naintenance which could result in nozzle blockage using an inspection of the nozzle or an air or smoke flow test to verify that the spray nozzles are not obstructed and that flow will be provided when required. The frequency is adequate to detect degradation in performance due to the passive nozzle design and its normally dry state and has been shown to be acceptable through operating, experience.

REFERENCES 1. USAR, Section 6.2.1.1.5.

2. .... ......e . .. ........

PERRY - UNIT 1 B 3.6-47 Revision No. 3

RHR Suppression Pool Cooling System B 3.6.2.3 BAS ES SURVEILLANCE SR 3.6.2.3.2 REQUIREMENTS (continued) Verifying each RHR pump develops a flow rate z 7100 gpm with flow through the associated heat exchanger' to the suppression pool, ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASME Section XI (Ref. 2). This test confirms one point on the pump design curve, and the results are indicative of cverall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.

REFERENCES 1. USAR., Section 6.2. * *..7_

2. 1 pd Pr-ess-r e aes1-lG A.d, S t4l PERRY - UNIT 1 B 3.6-82 Revision No. 1