ML043090576

Transmittal of Surry Power Station Units 1 and 2 Proposed Technical Specifications Change Request Relocation of Inservice Inspection and Testing Requirements

ML043090576
Person / Time
Site:	Surry
Issue date:	11/04/2004
From:	Hartz L Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
04-666
Download: ML043090576 (35)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 November 4, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Serial No.04-666 NL&OS/ETS RO Docket Nos. 50-280 License Nos. DPR-32 50-281 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE REQUEST RELOCATION OF INSERVICE INSPECTION AND TESTING REQUIREMENTS Pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests an amendment to Facility Operating License Numbers DPR-32 and DPR-37 in the form of changes to the Technical Specifications for Surry Power Station Units 1 and 2. The proposed changes will relocate the inservice testing requirements, remove the inservice inspection requirements, and establish a Bases Control Program consistent with Improved Technical Specifications. A discussion of the proposed changes is included in, Marked-up pages that identify the proposed changes and the Technical Specification pages that incorporate the proposed changes are provided in Attachments 2 and 3, respectively. Technical Specification Bases changes associated with the proposed changes are provided for information only. The Technical Specification Bases changes will be implemented following NRC approval of the license amendment.

The proposed changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee.

In accordance with the requirements of 10 CFR 50.92, the enclosed application is judged to involve no significant hazards. In addition, the proposed change has been determined to qualify for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(~)(9). The basis for these determinations is also included in.

Pursuant to 10 CFR 50.55a(f)(5)(ii) requirements, to avoid a conflict between inservice testing requirements stated in the Technical Specifications and the approved lnservice Testing Programs, Dominion requests approval of the proposed changes by May of 2005. The approved Technical Specifications will be implemented within 30 days of approval by the NRC staff.

Serial No.04-666 Docket Nos. 50-280/281 Page 2 of 3 Should you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Very truly yours, W

Leslie N. Hartz Vice President - Nuclear Engineering Attachments Commitments made in this letter: None cc:

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, Virginia 2321 8 Mr. N. P. Garrett NRC Senior Resident Inspector Surry Power Station Mr. S. R. Monarque NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8-HI2 Rockville, Maryland 20852

Serial No.04-666 Docket Nos. 50-280/281 Page 3 of 3 Technical Specification Change Inservice Testing and Bases Control Program COMMONWEALTH OF VIRGINIA

)

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz who is Vice President - Nuclear Engineering of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of her knowledge and belief.

774 Acknowledged before me this fl day of

,2004.

My Commission Expires:

(SEAL)

Proposed Technical Specification Changes For lnservice Testing and Bases Control Program Discussion of Change Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

DISCUSSION OF CHANGE Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests changes to the Technical Specifications (TS) for Surry Power Station Units 1 and 2.

The proposed changes will: 1) delete TS 4.0.5, which includes the Inservice Inspection (ISI) and lnservice Testing (IST) requirements; 2) relocate the IST requirements to the administrative section of TS as a program and revise the associated TSs to reference the IST program instead of TS 4.0.5; 3) deletes the inservice inspection surveillance requirements and individual TS references to the IS1 program since these requirements are included in 10 CFR 50.55a; and, 4) add a TS Bases Control Program to the Administrative Controls section of the TS.

The changes associated with the IST program are required pursuant to 10 CFR 50.55.a(f)(5)(ii) to eliminate an inconsistency between the TS and the approved ASME Code.

Furthermore, these IS1 and IST changes are consistent with NUREG-1 431, (Improved) Standard Technical Specifications for Westinghouse Plants. The associated Bases are also being revised and are included for information.

The proposed changes have been reviewed, and it has been determined that no significant hazards consideration exists as defined in 10 CFR 50.92. In addition, it has been determined that the change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(~)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

BACKGROUND AND DISCUSSION OF CHANGE lnservice Inspection and Testinq Proqrams TS 4.0.5 establishes the surveillance requirements for inservice inspection and testing of ASME Class 1, 2 and 3 components for Surry Power Station Units 1 and 2.

Regarding inservice testing (IST), TS 4.0.5.a currently states that:

lnservice Testing of ASME Code Class 1, 2, and 3 pumps and valves 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, Section 50.55a(f), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(f)(6)(i).

The regulations in 10 CFR 50.55a(f)(4) establish the effective Code edition and addenda to be used by licensees for performing inservice testing of pumps and valves.

Pursuant to 10 CFR 50a55a(f)(4)(ii), Virginia Electric and Power Company (Dominion) submitted the fourth interval IST programs for Surry Power Station Units 1 and 2 to the Page 1 of 8

NRC in a letter June 25, 2003 (Serial No.03-354). The IST Programs for the fourth interval were updated to comply with the appropriate revisions of the ASME Code for Operation and Maintenance of Nuclear Power Plants and included the 1998 Edition, the 1999 Addenda and the 2000 Addenda as the new Code of Record for performing IST at Surry Units land 2. As a consequence, the TS 4.0.5 reference to Section XI of the ASME Code results in a reference to a deleted portion of the Code.

According to 10 CFR 50.55a(f)(5)(ii), If a revised inservice test program for a facility conflicts with the technical specification for the facility, the licensee shall apply to the Commission for amendment of the technical specifications to conform the technical specification to the revised program. The licensee shall submit this application, as specified in 550.4, at least 6 months before the start of the period during which the provisions become applicable, as determined by paragraph (f)(4) of this section.

Since TS 4.0.5 and several Technical Specifications reference ASME Section XI for the IST requirements for pumps and valves, the Surry Units 1 and 2 TS require revision to change the IST code references from ASME Section XI to the ASME Code for Operation and Maintenance of Nuclear Power Plants.

For consistency with the Improved Technical Specifications (ITS), TS 4.0.5 will be relocated as an IST Program in the Administrative Section of TSs. In addition, those Technical Specifications that reference TS 4.0.5 will be changed to reference the new IST program.

Regarding inservice inspection (ISI), TS 4.0.5.a currently states that:

lnservice inspection of ASME Code Class 1, 2, and 3 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, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)( i).

When the ITS were developed the Inservice inspection requirements were removed since they were identical to the 10 CFR 50.55a requirements and were therefore, considered redundant. Thus, the IS1 requirements of TS 4.0.5 are being deleted, consistent with Improved Technical Specifications.

TS Bases Control Proqram The TS changes that are approved by the NRC Consolidated Line Item Improvement Program (CLIIP) often include the requirement for a TS Bases Control Program as a condition for implementing a particular Technical Specification Task Force (TSTF) traveler (e.g., TSTF 358 associated with missed surveillances, TSTF-359 associated with mode changes). As it is anticipated that future CLllP opportunities will invoke a similar requirement for a TS Bases Control Program and for consistency with ITS and Page 2 of 8

existing station administrative procedures, it is desirable to include such a program in Surrys custom Technical Specifications.

SPECIFIC CHANGES The following changes are proposed to move the IST requirements to a program in the Administrative Section of TS consistent with Improved Technical Specifications:

0 TS 4.0.5 is being relocated to Section 6 as an Inservice Testing Program and revised to include the appropriate ASME Code reference. In addition, the Inservice Inspection requirements in TS 4.0.5 are being deleted consistent with Improved Technical Specifications.

Delete: TS 4.0.5 and the associated Bases Insert: replace with the following IST Program in Section 6.

6.4.1 lnservice Testinq Proqram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

1. Testing frequencies specified in the ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda as follows:

ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda terminology for inservice testina activities Quarterly or every 3 months Yearly or annually Biennially or every 2 years Every 4 years Every 5 years Every 8 years Every 10 years Once per fuel cycle (18 months)

Every cold shutdown Every refueling outage Required Frequencies for performing inservice testinq activities At least once per 92 days At least once per 366 days At least once per 731 days At least once per 1461 days At least once per 1827 days At least once per 2922 days At least once per 3653 days At least once per 549 days Every cold shutdown Every refueling outage

2. The provisions of TS 4.0.2 are applicable to the above required Frequencies for performing inservice testing activities; Page 3 of 8

3. The provisions of TS 4.0.3 are applicable to inservice testing activities; and

4. Nothing in the ASME Code for Operation and Maintenance of Nuclear Power Plants shall be construed to supersede the requirements of any TS.

The frequencies identified in the table above are those identified in the ASME Code for Operation and Maintenance of Nuclear Power Plants.

The following TS are revised to refer to the Inservice Testing Program in place of TS 4.0.5:

J TS 4.1.B.1 - Pressuirzer PORVs and Block Valves J TS Table 4.1 -2A Item 4 - Pressurizer Safety Valves J TS Table 4.1 -2A Item 5 - Main Steam Safety Valves J TS 4.5.A.1 - Containment Spray pumps J TS 4.5.A.2 - Containment Spray valves J TS 4.5.B.1 - Recirculation Spray pumps J TS 4.5.8.1 - Recirculation Spray valves J TS 4.8.A.2.a - AFW valves J TS 4.8.A.3.a - AFW pumps J TS 4.8.A.5.b - AFW cross connect valves J TS 4.1 1.C.1 - Safey Injection Subsystem low head pumps J TS 4.1 1.C.2 - Safey Injection Subsystem charging pumps J TS 4.1 1.C.3 - Safety Injection Subsystem valves The following TS are revised to eliminate reference to ASME Section XI. Section XI is no longer the appropriate ASME Code of reference for the Inservice Testing Program.

J TS 4.8.B - Acceptance Criteria The following TS are revised due to the removal of the IS1 program requirements from the TS:

J TS 4.2.A - Augmented Inspections J TS Table 4.2 o Item 2.1.1 both examination frequency and frequency Item 2.1.2 both examination requirements and frequency o Item 2.1.3 both examination requirements and frequency J TS 4.17 Specification - SnubbersSection XI continues to be referenced in TS 4.2 as it relates to the examination methods and frequency of the augmented inspections.

Page 4 of 8

0 Delete Section 4.3, ASME Code Class 1, 2, and 3 System Pressure Tests and the associated Bases. Consistent with the removal of the IS1 requirements from TS, the pressure testing requirements are being removed from Technical Specifications.

Establish a Bases Control Program 6.4.J Technical Specifications (TS) Bases Control Proqram This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

2. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

a) a change in the TS incorporated in the license; or b) a change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

3. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

4. Proposed changes that meet the criteria of Specification 6.4.J.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

The title of Section 6.4 is being changed from Unit Operating Procedures to Unit Operating Procedures and Programs.

0 The following Bases Sections are being revised for consistency with the changes.

The Bases changes are provided for information only.

J Delete the Bases for TS 4.0.5 (pages 4.0-7&8)

J Bases 4.2 Item 2.1 and 2.2 (page 4.2-2)

J Bases 4.8 (page 4.8-3)

J Bases 4.1 1 (page 4.1 1-4)

Identified typos on the TS pages affected by this change are being corrected.

Page 5 of 8

SAFETY SIGNIFICANCE The proposed changes are administrative in nature, in that the changes only relocate and update the IST program, do not eliminate any tests (IST) or inspections (ISI), and establish a Technical Specification Bases Control Program.

The changes are consistent with Improved Techncial Specfications. These proposed changes will eliminate the ASME Code inconsistency between the IST program and the TS as required by 10 CFR 50.55a(f)(5)(ii). These administrative type changes to the TS have no impact on public health and safety.

NO SIGNIFICANT HAZARDS CONSIDERATION Virginia Electric and Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed Technical Specifications change to include an IST and Technical Specifications Bases Control Program for the Surry Units 1 and 2. We have determined that a significant hazards consideration is not involved as discussed below:

The proposed changes are administrative in nature. In the Federal Register, Vol. 51, No. 44, dated March 6. 1986, Rules and Regulations, the NRC provided guidance for the determination of significant hazards considerations. Under item (e) regarding examples of Technical Specifications amendments that are considered not likely to involve significant hazards consideration, the following example was listed:

(i)

A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

Technical issues concerning the Code were resolved previously through the NRC endorsement process which updates 10 CFR 50.55a(b)(2). 10 CFR 50.55a(f)(5)(ii),

states that if a revised inservice testing program for a facility conflicts with the technical specification for the facility, the licensee shall apply to the Commission for amendment of the technical specifications to conform the technical specification to the revised program. The proposed changes to the Technical Specifications (TS) for Surry Power Station Units 1 and 2 will: 1) delete TS 4.0.5 IS1 and IST requirements; 2) relocate the IST surveillance requirements to the administrative section of TS as a program and revise the associated TS to reference the IST program instead of TS 4.0.5; 3) delete the inservice inspection surveillance requirements and individual TS references to the IS1 program since these requirements are included in 10 CFR 50.55a; and 4) add a TS Bases Control Program to the Administrative Controls section of the TS. The changes associated with IST program are required pursuant to 10 CFR 50.55.a(f)(5)(ii) to eliminate an inconsistency between the TS and the approved ASME Code.

Furthermore, the IS1 and IST changes are consistent with NUREG-1431, Standard Technical Specifications for Westinghouse Plants.

Page 6 of 8

Criterion 1 - Operation of Surry Units 1 and 2 in accordance with the proposed Technical Specifications change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change is administrative in nature, and station operations are not being affected. The ASME Code requirements are established, reviewed and approved by ASME, the industry and ultimately endorsed by the NRC for inclusion into 10 CFR 50.55a.

Updates to the ASME Code reflect advances in technology and consider information obtained from plant operating experience to provide enhanced inspection and testing. Thus, the proposed change only modifies TS to appropriately reference the recently NRC approved Inservice Testing Program for the fourth interval at Surry Power Station. Consequently, the probability or consequences of an accident previously evaluated are not increased.

Criterion 2 - The proposed Technical Specifications change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

As noted above, the proposed change is administrative in nature, and no new accident precursors are being introduced.

Since the inservice testing will continue to be performed in accordance with an NRC approved program, adequate assurance is provided to ensure the safety-related pumps and valves would operate as required. No new testing is required that could create a new or different type of accident.

Consequently, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3 - The proposed Technical Specifications change does not involve a significant reduction in a margin of safety.

Performing inservice testing of pumps and valves to the NRC approved program for the fourth interval at Surry Power Station provides adequate assurance that the safety-related pumps and valves will continue to perform their intended safety function. This is an administrative change in nature and as such does not involve a significant reduction in the margin of safety.

ENVIRONMENTAL ASSESSMENT This amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(~)(9) as follows:

(i)

The amendment involves no significant hazards consideration.

As described above, the proposed change involves no significant hazards consideration.

Page 7 of 8

(ii)

There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change does not involve the installation of any new equipment, or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite. This change only establishes programs to address Inservice Testing activities and the Bases of Surrys Technical Specification.

Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii)

There is no significant increase in individual or cumulative occupation radiation exposure.

The proposed change does not involve plant physical changes, or introduce any new mode of plant operation. This change only establishes programs to address Inservice Testing activities and the Bases of Surrys Technical Specification.

Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above, Dominion concludes that the proposed changes meet the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.22 relative to requiring a specific environmental assessment by the Commission.

Page 8 of 8 Proposed Technical Specification Changes For lnservice Testing and Bases Control Program Mark-Up of Proposed Changes Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Do m i n i o n)

TS ii w

TECHNICAL SPECIFICATION TABLE OF CONTENTS SECTION 3.15 3.16 3.17 3.18 3.19 3.20 3.21 3.22 3.23 TITLE DELETED EMERGENCY POWER SYSTEM LOOP STOP VALVE OPERATION MOVABLE INCORE INSTRUMENTATION MAIN CONTROL ROOM BOTTLED AIR SYSTEM SHOCK SUPPRESSORS (SNUBBERS)

DELETED AUXILIARY VENTILATION EXHAUST HLTER TRAINS CONTROL AND RELAY ROOM VENTILATION SUPPLY FILTER TRAINS 4.0 SURVEILLANCE REQUIREMENTS 4.1 OPERATIONAL SAFETY REVIEW PAGE TS 3.16-1 TS 3.17-1 TS 3.18-1 TS 3.19-1 TS 3.20-1 I

TS 3.22-1 TS 3.23-1 TS 4.0-1 TS 4.1-1 TS 4.2-1 4.4 4.5 4.6 4.7 4.8 4.9 4.10 4.1 1 4.12 4.13 4.14 CONTAINMENT TESTS SPRAY SYSTEMS TESTS EMERGENCY POWER SYSTEM PERIODIC TESTING MAIN STEAM LINE TRIP VALVES AUXILIARY FEEDWATER SYSTEM RADIOACTIVE GAS STORAGE MONITORING SYSTEM REACTIVITY ANOMALIES SAFETY INJECTION SYSTEM TESTS VENTILATION FILTER TESTS DELETED DELETED TS 4.4-1 TS 4.5-1 TS 4.6-1 TS 4.7-1 TS 4.8-1 TS 4.9-1 TS 4.10-1 TS 4.11-1 TS 4.12-1 Amendment Nos. -17

TS iii TECHNICAL SPECIFICATION TABLE OF CONTENTS SECTION 4.15 4.16 4.17 4.18 4.19 4.20 TITLE AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT LEAKAGE TESTING OF MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES SHOCK SUPPRESSORS (SNUBBERS)

DELETED STEAM GENERATOR INSERVICE INSPECTION CONTROL ROOM AIR FILTRATION SYSTEM DESIGN FEATURES 5.1 SITE 5.2 CONTAINMENT 5.3 REACTOR 5.4 FUEL STORAGE ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW 6.2 6.3 6.4 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED UNIT OPERATING PROCEDUREy,4A/i> BZoC7R4 HS ) -

u 6.5 STATION OPERATING RECORDS 6.6 STATION REPORTING REQUIREMENTS 6.7 ENVIRONMENTAL QUALIFICATIONS 6.8 PROCESS CONTROL PROGRAM AND OFFSITE DOSE CALCULATION MANUAL PAGE TS 4.15-1 TS 4.16-1 TS 4.17-1 I

TS 4.19-1 TS 4.20-1 TS 5.1-1 TS 5.1-1 TS 5.2-1 TS 5.3-1 TS 5.4-1 TS 6.1-1 TS 6.1-1 TS 6.2-1 TS 6.3-1 TS 6.4-1 1

TS 6.5-1 TS 6.6-1 TS 6.7-1 TS 6.8-1 Amendment Nos. 2 G 4 ~ e U l A

TS 4.0-2 rmed in accor Amendment Nos. Z%kd&KL

TS 4.0-3 f i r % l3J-r

+s pLoG&eH '."

Seen -3 u' 6. 4 Amendment Nos. 1-

83.3293 TS 4*0-7 %

Under the provisions of this specification, the applicable surveillance requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

Exceptions to Specification 4.0.4 allow performance of surveillance requirements associated with a Limiting Condition for Operation after entry into the applicable operational condition.

When a shutdown is required to comply with Action Statement requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower condition of operation.

inservice testing of ection XI of the A portions of the abo rements has b pecification includ remove an 7

)

?

7 Amendment Nos. 45%d-H+

TS 4.0-8

-4%&!+3 before a device that is Amendment Nos. 155zmH3'4

TS 4.1-1 4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Obiective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Specification A. Calibration, testing, and checking of instrumentation channels and interlocks shall be performed as detailed in Tables 4.1-1,4.1-A, and 4.1-2.

B. Equipment tests shall be performed as

1. In addition to the requirements shall be demonstrated OPERABLE by:

a.

Performing a complete cycle of each PORV with the reactor coolant average temperature >350"F once per 18 months.

b.

Performing a complete cycle of the solenoid air control valve and check valves on the air accumulators in the PORV control system once per 18 months.

c.

Operating each block valve through one complete cycle of travel at least once per 92 days. This surveillance is not required if the block valve is closed in accordance with 3.1.6.a, b, or c.

d.

Verifying that the pressure in the PORV backup air supply is greater than the surveillance limit at least once per 92 days.

e.

Performing functional testing and calibration of the PORV backup air supply instrumentation and alarm setpoints at least once per 18 months.

Amendment Nos. 2 3 k d 4 3 4

DESCRIPTION

1. Control Rod Assemblies

2.

3.

4.

5.

6.

7.

8.

9.

Control Rod Assemblies Refueling Water Chemical Addition Tank Pressurizer Safety Valves Main Steam Safety Valves Containment Isolation Trip Refueling System Interlocks Service Water System Deleted TABLE 4.1 -2A MINIMUM FREQUENCY FOR EQUIPMENT TESTS TEST Rod drop times of all full length rods at hot conditions Partial movement of all rods Functional Setpoint Setpoint

• Functional

• Functional

• Functional FSAR SECTION FREQUENCY REFERENCE Prior to reactor criticality:

7

a. For all rods following each removal of the reactor vessel head

b. For specially affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. Once per 18 months Quarterly 7

Once per 18 months 6

Prior to refueling 9.12 Once per 18 months 9.9

10. Primary System Leakage

• Evaluate Daily 1 1. Diesel Fuel Supply

• Fuel Inventory 5 dayslweek

12. Deleted

13. Main Steam Line Trip Valves Functional (Full Closure)

Before each startup (TS 4.7)

The provisions of Specification 4.0.4.

are not applicable I

This page was oblished electronically for use on the MIND system. Differences between this page and a age from the hardco y version of the Technical Specifications are differences in (appearance on!.

Such differences are intentional and are the result of mana ing an electronic master of !he station's Technica~Specifications. The accuraw of the content of the MIND version of the echnical Soecifications has been confirmed hv Confiauratinn kmamment.

4 8.5 10

TS 4.2-1 Aa%e+

4.2 AUGMENTED INSPECTIONS Applicability Applies to inservice inspections which augment those required by ASME Section XI.

Objective To provide the additional assurance necessary for the continued integrity of important components involved in safety and plant operation.

Specifications A. Inspections shall be performed as specified in TS. Table 4.2-1. Nondestructive examination techniques and acceptance criteria shall be in compliance with the requirements of A %

B. The normal inspection interval is 10 years.

C. Detailed records of each inspection shall be maintained to allow a continuing evaluation and comparison with future inspections.

Bases The inspection program for ASME Section XI of the ASME Boiler and Pressure Vessel Code limits its inspection to ASME Code Class 1, 2, and 3 components and supports.

Certain components, under Miscellaneous Inspections in this section, were added because of no corresponding code requirement. This added requirement provides the inspection necessary to insure the continued integrity of these components.

i Item 1.4 The low pressure turbine rotor blades are normally inspected concurrent with the disk and hub inspections. The disk and hub inspection frequency is based on existing crack size, crack growth rate, and system operating conditions. ASME Section XI does not provide specific examination requirements or acceptance criteria for turbine rotor inspections.

Procedures and acceptance criteria for turbine rotor inspections are consistent with general industry practices.

Amendment Nos. 1444+l?!,

il R

Z 0

?

TABLE 4.2-1 (continued)

SECTION B. SENSITIZED STAINLESS STEEL Required Examination Item No.

2.1.1 Class 1 circumferential,

longitudinal, branch pipe connection, and socket welds 2.1.2 Class 2 circumferen ti a1,

longitudinal, branch pipe connection, and socket welds 2.1.3 Class 1 and Class 2 sensitized stainless steel pieces Required Methods Interval Inspection Remarks Examination 1 0-Year As required by The welds examined by volumetric or surface techniques shall be conducted at three times the frequency required A minimum of 5% of the welds shall be examined once per 18 months. At least 75% of the total population of welds shall be examined each interval. The same welds may be selected in subse-quent intervals for examination. See Note 1.

The welds examined by A minimum of 2.5% of the welds shall be examined once per 18 months. At least 22.5% of the total population of welds shall be examined each interval.

The same welds may be selected in sub-volumetric or surface techniques shall be conducted at three times the frequency required sequent intervals for examination. See Note 1.

by W. 4.6.5 In addition to the Code required exami-required by nations the affected piping shall be visually (VT-2) examined during the flushing requirements of T.S. Tables 4.1 -3A and 4.1 -3B.

This page was ublished electronically for use on the MIND system. Differences between this page and a age from the hardco y version of the Technical Specifications are differences in appearance on!.

Such differences are intentional and are the result of mana ing an electronic master of !he station's Technicaf Specifications. The accuracy of the content of the MIND version of the ethnical Specifications has been confirmed by Configuratlon flanagement.

TS 4.2-2 Sensitized stainless steel augmented inspections were added to assure piping integrity of this classification.

The examinations required by k%en@tilize the periodically updated ASME Section XI Boiler and Pressure Vessel Code

.. -nrt

+r- &

natio$?The surface and volumetric examinations required by*

ite t three times the frequency required by the Code in an interval. In addition to the Code required pressure testing, visual examinations will be conducted, while the piping is pressurized by the procedures defined in Tables 4.1 -3A & B of Technical Specification 4.1, concerning flushing of sensitized stainless steel piping. Weld selection criteria are modified from the Code for Class 1 welds, since stress level information as correlated to weld location is unavailable for Surry.

1 I

Item 2.2. I The sensitized stainless steel located in the containment and recirculation spray rings in the overhead of containment are classified ASME Class 2 components. These components are currently exempted by ASME Section XI from surface and volumetric examination requirements. As such, an augmented program will remain in place requiring visual (VT-1) examination of these components for evidence of cracking. Additionally, sections of the piping will be examined by liquid penetrant inspection when the piping is visually nspected.

Amendment Nos.

efined as the state of rmal operation aft for ASME Cod A. Inservice ins rmed in accord w

Amendment Nos. 1

TS 4.3-2 1+0590t-Amendment Nos. 1

TS 4.5-1 4.5 SPRAY SYSTEMS TESTS Applicability Applies to the testing of the Spray Systems.

Objective To verify that the Spray Systems will respond promptly and perform their design function, if required.

Specification A. Each containment spray subsystem shall be demonstrated OPERABLE:

1. By verifying, that on recirculation flow, each containment spray pump performs satisfactorily when tested in accordance with @Y??E2z&*-2

/

/?byrn&

L-

2. By verifying that each motor-operated valve in the contazment 6 flow performs satisfactorily when tested in accordance wit

3. By verifying each spray nozzle is unobstructed following maintenance which could cause nozzle blockage.

4. Coincident with the containment spray pump test described in Specification 4.5.A. 1, by verifying that no particulate material clogs the test spray nozzles in the refueling water storage tank.

B. Each recirculation spray subsystem shall be demonstrated OPERABLE:

1. By verifying each recirculation spray pump performs satisfactorily when tested in I

Amendment Nos. B%imlTE

TS 4.5-2

&Hf+m

2. By verifying that each motor-operated valve in the recirculation spray flow paths performs satisfactorily when tested in accordance wi

3. By verifying each spray nozzle is unobstructed following maintenance which could cause nozzle blockage.

C. Each weight-loaded check valve in the containment spray and outside containment recirculation spray subsystems shall be demonstrated OPERABLE once per 18 months by cycling the valve one complete cycle of full travel and verifying that each valve opens when the discharge line of the pump is pressurized with air and seats when a vacuum is applied.

D. A visual inspection of the containment sump and the inside containment recirculation spray pump wells and the engineered safeguards suction inlets shall be performed once per 18 months and/or after major maintenance activities in the containment. The inspection should verify that the containment sump and pump wells are free of debris that could degrade system operation and that the sump components (i,e., trash racks, screens) are properly installed and show no sign of structural distress or excessive corrosion.

Amendment Nos. 2Ghm&HY

TS 4.8-1 03-07-94 4.8 AUXILIARY FEEDWATER SYSTEM Applicability Applies to the periodic testing requirements of the Auxiliary Feedwater System.

m c t i v e To verify the operability of the auxiliary feedwater pumps.

Specification A. Tests and Frequencies

1. At least once per 31 days:

a. Verify that the Auxiliary Feedwater System manual, power operated, and automatic valves in each flow path are in the correct position. This verification includes valves that are not locked, sealed, or otherwise secured in position, valves in the the cross-connect from the opposite unit and valves in the steam supply paths to the turbine driven auxiliary feedwater pump.

2. At least once per 92 days:

a. Verify that each motor-operated valve in the auxiliary feedwater flow paths, including the cross-connect from the opposite unit, performs satisfactorily when tested in accordance wit

3. At least once per 9

a. Verify that the ctorily when tested in he provisions of Specification 4.0.4 are accordance wit not applicable fo Amendment Nos. 190 and 190

TS 4.8-2 4iw3+9+

7 4a. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to Reactor Coolant System temperature and pressure exceeding 350°F and 450 psig, respectively, the motor driven auxiliary feedwater pumps shall be flow tested from the 110,000 gallon above ground Emergency Condensate Storage Tank to the steam generators.

4b. Wthin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after achieving reactor criticality, the steam turbine driven auxiliary feedwater pump shall be flow tested from the 110,000 gallon above ground Emergency Condensate Storage Tank to the steam generators. The provisions of Specification 4.0.4 are not applicable.

5. During periods of reactor shutdown with the opposite units Reactor Coolant System temperature and pressure greater than 350°F and 450 psig, respectively:

a. Continue to verify that the motor driven auxiliary feedwater pump perform satisfactorily when tested at the frequency defined in Specification 4.8.A.3.

b. Verify that each motor-operated valve in the auxiliary feedwater cross-connect flow path for the opposite unit performs satisfactorily when tested in accordance B. Acceptance Criteria The pump and valve tests, except the system flow test, shall be considered satisfactory if they meet the The system flow tests during unit startup from COLD SHUTDOWN or REFUELING SHUTDOWN shall be considered satisfactory if the control board indication demonstrates that flow paths exist to each steam generator.

Inservice Testing Program acceptance criteria.

Amendment Nos. +9hn+W-

TS 4.8-3 03-07-94 Basis The correct alignment for manual, power operated, and automatic valves in the Auxiliary Feedwater System steam and water flow paths, including the cross-connect flow path, will provide assurance that the proper flow paths exist for system operation. This position check does not include: 1) valves that are locked, sealed or otherwise secured in position since they are verified to be in their correct position prior to locking, sealing or otherwise securing; 2) vent, drain or relief valves on those flow paths; and, 3) those valves that cannot be inadvertently misaligned such as check valves. This surveillance does not require any testing or valve manipulation. It involves verification that those valves capable of being mispositioned are in the correct position.

The auxiliary feedwater pump will be tested periodically in accordance

&ctie&H to demonstrate operability. The pumps are flow tested on recirc 110,000 gallon Emergency Condensate Storage Tank. Valves in the flow pat generators and cross-connect flow path are tested periodically in accordanc

&Scp

• Z X - 7 5 7-9 r-d -

L-Z The auxiliary feedwater pumps are capable of supplying feedwater to the opposite units steam generators. For a main steam line break or fire event in the Main Steam Valve House, one of the opposite units auxiliary feedwater pumps is required to supply feedwater to mitigate the consequences of those accidents. Therefore, when considering a single failure, both motor driven auxiliary feedwater pumps are required to be OPERABLE* during shutdown to support the opposite unit if the Reactor Coolant System temperature or pressure of the opposite unit is greater than 350°F and 450 psig, respectively. Thus, to establish operability* the motor driven auxiliary feedwater pumps will continue to be tested quarterly on the same STAGGERED TEST BASIS when the unit is shutdown to support the opposite unit. The turbine driven pump is not required to be OPERABLE when the unit is shutdown and therefore, is not tested during periods of shutdown.

• excluding automatic initiation instrumentation Amendment Nos. 190 and 190

TS 4.11-2

2.

At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution.

a.

This surveillance is not required when the volume increase makeup source is the RWST.

C.

Each Safety Injection Subsystem shall be demonstrated OPERABLE:

1.

By verifying, that on recirculation flow, each low head safety injection pump performs satisfactorily when tested in accordance with By verifying that each charging pump performs saitisfactor accordance wit By verifying that each motor-operated valve in the sa performs satisfactorily when tested in accordance with

2.

3.

4.

Prior to POWER OPERATION by:

a.

Verifying that the following motor operated valves are blocked open by de-energizing AC power to the valves motor operator and tagging the breaker in the off position:

Unit 1 Unit 2 MOV-1890C MOV-2890C:

b.

Verifying that the following motor operated valves are blocked closed by de-energizing AC power to the valves motor operator and the breaker is locked, sealed or otherwise secured in the off position:

Unit 1 Unit '2 MOV-1869A MOV-2869A MOV-1869B MOV-2869B MOV-1890A MOV-2890A MOV-1890B MOV-2890B Amendment Nos. 199 and 199

TS 4.11-4 The system tests demonstrate proper automatic operation of the Safety Injection System. A test signal is applied to initiate automatic operation action and verification is made that the components receive the safety injection signal in the proper sequence. The test may be performed with the pumps blocked from starting.

The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.

During reactor operation, the instrumentation which is depended on to initiate safety injection is checked periodically, and the initiating circuits are tested in accordance with Specification 4.1. In addition, the active components (pumps and valves) are to be periodically tested to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval is determined in accordance with

. The accumulators are a Dassive I

safeguard.

References UFSAR Section 6.2, Safety Injection System Amendment Nos. 1

TS 4.17-1 83-f-2-43y 4.17 SHOCK SUPPRESSORS (SNUBBERS)

Applicability Applies to all hydraulic and mechanical shock suppressors (snubbers) which are required to protect the Reactor Coolant System and other safety-related systems. Snubbers excluded from this inspection are those installed on non-safety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

Obiective To specify the minimum frequency and type of surveillance to be applied to the hydraulic and mechanical snubbers required to protect the Reactor Coolant System and other safety-related systems.

Specification Each snubber shall be demonstrated OPERABLE by inservice inspection program and the requirements b.

As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

A. Visual Inspections

1. Snubbers are categorized as inaccessible or accessible during reactor operation.

Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 4.17-1. The visual inspection interval of each category of snubber shall be determined based upon the criteria provided in Table 4.17-1.

Amendment Nos. WEfEff%