ML070510048

Proposed Revision to Technical Specifications (LBDCR 05-MP2-008) Pressurizer Power Operated Relief Valves (Porvs) Testing

ML070510048
Person / Time
Site:	Millstone
Issue date:	02/16/2007
From:	Gerald Bichof Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0006
Download: ML070510048 (20)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginia 23060 Wet) Address: www.dom.com February 16, 2007 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.

07-0006 MPS LicNVDB RO Docket No.

50-336 License No.

DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP2-008)

PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) TESTING Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License DPR-65 for Millstone Power Station Unit 2 (MPS2). The enclosed license amendment request proposes to revise Technical Specification 314.4.3, "Reactor Coolant System, Relief Valves." The proposed change modifies the method of testing of pressurizer Power Operated Relief Valves (PORVs).

The proposed amendment does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92. The Site Operations Review Committee has reviewed and concurred with this determination. contains description of the proposed Technical Specification (TS) change and the Significant Hazards Consideration. Attachment 2 contains the TS marked-up pages. Attachment 3 contains the marked-up pages of the TS Bases for information only. MPS2 TS Bases are controlled in accordance with TS Section 6.23, "Technical Specification Bases Control Program."

Issuance of this amendment is requested no later than December 30, 2007, with the amendment to be implemented within 90 days of issuance.

In accordance with 10 CFR 50.91 (b), a copy of this license amendment request is being provided to the State of Connecticut.

If you have any questions or require additional information, please contact Mr. Paul R.

Willoughby at (804) 273-3572.

Very truly yours, Gerald T. Bischof

(/

Vice President - Nuclear Engineering

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 2 of 3 Attachments:

1. Evaluation of Proposed License Amendment

2. Marked-Up TS Page

3. Marked-Up TS Bases Page, for information only Commitments made in this letter: None.

cc:

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11 555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 061 06-5 127

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 3 of 3 COMMONWEALTH OF VIRGINIA

)

1 COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President - Nuclear Engineering, of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he 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 his knowledge and belief.

Acknowledged before me this /b" day of & uu

, 2007.

My Commission Expires:

&UT

.Z ~ w.

(SEAL)

Serial No. 07-0006 Docket No. 50-336 ATTACHMENT I PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP2-0081 PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) TESTING EVALUATION OF PROPOSED LICENSE AMENDMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 1 of 10 EVALUATION OF PROPOSED LICENSE AMENDMENT

3.0 BACKGROUND

3.1 Description of the Pressurizer Power Operated Relief Valves 3.2 The Method For PORV Testing 3.3 Reason for the Proposed Amendment

4.0 TECHNICAL ANALYSIS

4.1 Details of the Proposed Amendment 4.2 Testing of Replacement PORVs 4.3 Other Affected TSs 4.4 Safety Summary

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirementslcriteria

7.0 REFERENCES

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 2 of 10

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests to amend Operating License DPR-65 for Millstone Power Station Unit 2 (MPS2).

This license amendment request proposes to revise Technical Specification (TS) 314.4.3, "Reactor Coolant System, Relief Valves." The proposed change eliminates an unnecessary level of detail regarding the method of testing of pressurizer Power Operated Relief Valves (PORVs).

2.0 PROPOSED CHANGE

Modify Surveillance Requirement (SR) 4.4.3.1~ by replacing the wording:

"Once per 18 months the PORVs shall be bench tested at conditions representative of MODES 3 or 4."

With "Once per 18 months by operating the PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4."

Bases Changes TS Bases Section 314.4.3, "Relief Valves," will also be updated to reflect the proposed TS changes. The TS Bases changes are provided for information only in. Changes to the Bases are controlled in accordance with the TS Bases Control Program (TS 6.23).

3.0 BACKGROUND

3.1 Description of the Pressurizer Power Operated Relief Valves Overpressure protection for the Reactor Coolant System (RCS) is provided by two PORVs and two ASME Code safety valves as described in the MPS2 Final Safety Analysis Report (FSAR) Section 4.3. The PORVs (2-RC-4021404) share a common tap from the pressurizer. The PORVs were replaced during the last refueling outage in the fall 2006 (2R17) with solenoid operated valves that fail closed on a loss of control power. Additionally, the new PORVs are credited as the pressurizer high point vents in accordance with NUREG-0737 and 10 CFR 50.46a. This permitted removal of the separate pressurizer high point vent system as part of the pressurizer and PORV replacement activities during 2R17. It was determined that this modification did not require prior NRC approval per 10 CFR 50.59. The two PORVs relieve sufficient pressure during abnormal transients to prevent opening of the RCS safety valves. The PORVs open when 2 out of 4 of the pressurizer

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 3 of 10 pressure safety channels exceed the high setpoint and reclose when 3 out of 4 channels drop below the setpoint. The PORVs are also used when the RCS is at a reduced pressure and temperature to provide for low temperature over pressure (LTOP) protection.

The new PORVs are environmentally and seismically qualified. The associated power supplies, electrical components and circuitry are safety related. This level of reliability is consistent with the requirements of NUREG-0737.

The new PORVs are direct-acting solenoid operated valves, whereas the old valves were indirect-acting pilot-operated solenoid valves. No new failure modes were introduced by the subject valve replacement.

A motor operated gate valve, PORV Block Valve, (2-RC-4031405) is provided upstream of each of the relief valves to permit isolation of the PORVs in the event of a failure or excessive leakage.

Each PORV also has an openlclose keyswitch to allow manual opening of the valve to reduce RCS pressure in an emergency. The PORVs receive an actuation signal from the reactor protection system (RPS) when two of four pressurizer pressure safety channels exceed their setpoint. The PORVs relieve to the quench tank. Temperature is monitored in the piping downstream of the relief valves to detect valve leakage or valve actuation.

The temperature elements in the pressurizer relief discharge line supply signals for indication and alarm in the control room. Additionally, the new PORVs are provided with position indication lights in the control room.

The PORVs are also used to provide LTOP protection. There is a bypass circuit around the contacts that open the PORVs on high pressure. The bypass circuit contains three sets of contacts. One set of contacts closes when the LTOP setpoint selector switch is selected to LOW. The next set of contacts closes when the pressurizer pressure is greater than 415 PSIA. The third set of contacts closes when the RCS temperature is less than 275" F. When all three contacts are closed, the respective PORV is energized to open.

If the RCS temperature subsequently increases above 275" F, the bypass circuitry is removed without regard to positioning of the LTOP setpoint selector switch.

An interlock prevents closing the motor operated block valves for the PORVs when the LTOP setpoint selector switch is selected to LOW, thus ensuring that overpressure protection is not isolated when the plant is in the LTOP protection mode.

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 4 of 10 3.2 Method for PORV Testing In accordance with existing SR 4.4.3.1.c, and in addition to the requirements of TS 4.0.5, the old PORVs were bench tested by a qualified laboratory once per 18 months at conditions representative of MODES 3 or 4. This testing also satisfied the testing requirement of the MPS2 In-Service Test Program as specified in SR 4.4.9.3.1.d.

3.3 Reason for the Proposed Amendment The proposed change eliminates an unnecessary level of detail regarding the method of testing of pressurizer Power Operated Relief Valves (PORVs). The replacement upgraded PORVs are welded in place, and in-situ testing is the preferred methodology for this testing. Therefore, the SR wording is being revised to eliminate reference to a specific method of testing (i.e., bench testing).

The proposed TS change provides greater flexibility in testing, which is necessary given that the replacement PORVs are welded in place.

4.0 TECHNICAL ANALYSIS

4.1 Details of the Proposed Amendment The PORVs were replaced during the fall 2006 refueling outage (2R17) as part of the project to replace the MPS2 pressurizer. The replacement PORVs continue to provide overpressure protection for the RCS by relieving pressure from the RCS through the pressurizer and blowing down to the quench tank, as before. The PORV minimum opening pressure and flow capacity remains the same.

The enclosed license amendment request proposes to revise SR 4.4.3.1.c by replacing the wording:

"Once per 18 months the PORVs shall be bench tested at conditions representative of MODES 3 or 4."

With "Once per 18 months by operating the PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4."

The proposed change continues to require testing of the PORVs at conditions representative of MODES 3 or 4. The current MPS2 TS was based on the guidance contained in GL 90-06 as approved in NRC Issuance Of Amendment (TAC NO. M89380) dated February 15, 1995. With the exception of removing the

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 5 of 10 reference to bench testing, there is no change in the current licensing basis associated with this change. The proposed change will continue to allow the option of bench testing at conditions representative of MODES 3 or 4 if the valves should need to be removed for vendor required maintenance. The remainder of SR 4.4.3.1 is unaffected by the proposed change.

4.2 Testing of Replacement PORVs Pre-Installation Stroke Testing:

Initial stroke testing to verify OPERABILITY for MODE 1, 2 and 3 during operating cycle 18 was accomplished at the vendor facility prior to shipment and installation in the plant.

Surveillance Stroke Testing Everv 18 Months:

For subsequent operating cycles, at least once per 18 months each PORV will be operated through one complete cycle of full travel at conditions representative of MODES 3 or 4. This will normally be performed in MODE 3 or 4 as the unit is descending in power to commence a refueling outage. This test will normally be a static test, whereby a PORV will be exposed to MODE 3 or 4 temperatures, the block valve closed, and the PORV tested to verify it strokes through one complete cycle of full travel. This testing will be performed prior to establishing conditions where the PORVs are used for LTOP.

This test satisfies the GL 90-06 requirements.

Testing in the manner described is also consistent with the guidance in NUREG-1482, "Guidelines for lnservice Testing at Nuclear Power Plants," Section 4.2.10, that describes the PORVs function during reactor startup and shutdown to protect the reactor vessel and reactor coolant system from LTOP conditions, and indicates the valves should be exercised before system conditions warrant vessel protection. The PORV testing requirements of TS 4.0.5 specified in SR 4.4.3.1 remain unchanged.

Post-Maintenance Retest:

If post-maintenance retest is warranted, the affected valve(s) will be stroked under ambient conditions while in Mode 5 or 6 or while defueled. The actual stroke time in the open and close direction will be measured, recorded and compared to the test results obtained during pre-installation testing to assess operability of the affected valve(s). The stroke time testing of these valves in the IST Program involves removal of the solenoid cover and the installation of current transducers on the coil power lead and on each of the position indication light leads. Current traces are recorded during the stroke test by the Teledyne Quik-Look diagnostic system, which is used to determine stroke times to the nearest millisecond. These data traces are also overlaid for comparison with baseline current traces from production testing. Use of this diagnostic system affords specific and trendable performance data beyond simple stroke timing or exercise testing. Installation of the current transducers is impracticable during conditions of startup and shutdown.

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 6 of 10 Performance of this testing following maintenance will enable a comprehensive assessment of the effect of maintenance on valve operation.

4.3 Other Affected TSs The proposed amendment does not impact any other sections of the MPS2 TSs.

4.4 Safety Summary The proposed change will ensure that testing of PORVs, in addition to the requirements of TS 4.0.5, continues to be performed at least once per 18 months at conditions representative of MODES 3 or 4. The proposed method of testing is consistent with the staff position outlined in GL 90-06 as approved in NRC Issuance Of Amendment (TAC NO. M89380) dated February 15, 1995. The proposed change eliminating unnecessary detail regarding the method of testing provides the same level of assurance of PORV OPERABILITY in the required MODES of operation as the current method of testing (bench testing). Therefore, the proposed change has no adverse effect on plant safety.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration The proposed amendment permits the method used for Power Operated Relief Valve (PORV) testing to be changed from bench testing at conditions representative of MODES 3 or 4 to operating each PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4. The testing frequency of once per 18 months remains unchanged. In accordance with 10 CFR 50.92, Dominion Nuclear Connecticut, Inc., (DNC) has reviewed the proposed change and has concluded that it does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised as detailed below.

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

Response: No.

The proposed change does not modify any plant equipment and does not impact any failure modes that could lead to an accident. Additionally, the proposed change has no effect on the consequence of any analyzed accident since the change does not affect the function of any equipment credited for accident mitigation. In-situ testing versus bench testing does not decrease the reliability of the PORVs. Based on this discussion, the proposed amendment

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 7 of 10 does not increase 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 change does not modify any plant equipment and there is no impact on the capability of existing equipment to perform its intended functions.

No system setpoints are being modified and no changes are being made to the method in which plant operations are conducted. No new failure modes are introduced by the proposed change. The proposed amendment does not introduce accident initiators or malfunctions that would cause a new or different kind of accident. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The TS change does not involve a significant reduction in a margin of safety because the acceptance criterion (i.e., demonstration of function by operation of the PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4) for the valve testing is the same. The proposed change does not affect any of the assumptions used in the accident analysis, nor does it affect any operability requirements for equipment important to plant safety. Therefore, the margin of safety is not impacted by the proposed amendment.

As described above, this License Amendment Request does not impact the probability of an accident previously evaluated, does not involve a significant increase in the consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident from any accident previously evaluated, and does not result in a significant reduction in a margin of safety.

Therefore, DNC has concluded that the proposed changes do not involve an SHC.

5.2 Applicable Regulatory Requirementstcriteria The requirements of General Design Criterion (GDC) 15 of Appendix A to Title 10 of the Code of Federal Regulations Part 50 (10 CFR Part 50) state that "the reactor coolant system and associated auxiliary, control and protection systems shall be

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 8 of 10 designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences."

The pressurizer, associated PORVs and safety valves are key components which assure compliance with GDC 15. The proposed change is administrative in nature and does not impact the frequency of testing or the acceptance standards used for evaluating OPERABILITY. As such, compliance with GDC 15 is maintained.

Additionally, the staff concluded, in GL 90-06, that the following actions should be taken to improve the reliability of PORVs and block valves:

1. Include PORVs and block valves within the scope of an operational quality assurance program that is in compliance with 10 CFR Part 50, Appendix B.

This program should include the following elements:

a. The addition of PORVs and block valves to the plant operational Quality Assurance List.

b. Implementation of a maintenancelrefurbishment program for PORVs and block valves that is based on the manufacturer's recommendations or guidelines and is implemented by trained plant maintenance personnel.

c. When replacement parts and spares, as well as completed components, are required for existing non-safety-grade PORVs and block valves (and associated control systems), it is the intent of GL 90-06 that these items may be procured in accordance with the original construction codes and standards.

2. PORVs, PORV control air systems and block valves should be covered by subsection IWV of Section XI of the ASME Code. PORV testing should be performed in Modes 3 or 4.

3. Modify the limiting conditions for operations for Modes 1, 2, and 3 as specified in Attachments A-I through A-3 of GL 90-06.

MPS2 is complying with all the requirements of GL 90-06 as documented in DNC responses to GL 90-06 dated December 21, 1990, March 21, 1991, August 20, 1992, January 1 I, 1993, and April 25, 1994. The proposed changes to TS are also consistent with the requirements of GL 90-06.

Additionally, the replacement PORVs serve the function as the pressurizer high point vents per 10 CFR 50.46a1 which permitted removal of the separate pressurizer high point vent system as part of the pressurizer and PORV

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 9 of 10 replacement activities during 2R17. It was determined that this modification did not require prior NRC approval per 10 CFR 50.59.

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.

6.0 ENVIRONMENTAL CONSIDERATION

DNC has determined that the proposed amendment would change requirements with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, or it would change inspection or surveillance requirements.

DNC has evaluated the proposed change and has determined that the 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 off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(~)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. Generic Letter 90-06, Resolution Of Generic lssue 70, "Power-Operated Relief Valve And Block Valve Reliability," And Generic lssue 94, "Additional Low-Temperature Overpressure Protection For Light-Water Reactors," Pursuant To 10 CFR 50.54(f), dated June 25, 1990.

2. E. J. Mroczka letter B13694 to the U. S. Nuclear Regulatory Commission, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos. 2 and 3, Generic Letter 90-06, Resolution of Generic lssue 70, 'Power-Operated Relief Valve and Block Valve Reliability,' and Generic lssue 94, 'Additional Low-Temperature Overpressure Protection for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f)," dated December 21, 1990.

3. E. J. Mroczka letter B13760 to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Nos. 2 and 3, Generic Letter 90-06, Resolution of Generic lssue 70, 'Power-Operated Relief Valve and Block Valve Reliability,' and Generic lssue 94, 'Additional Low-

Serial No. 07-0006 Docket No. 50-336 Power Operated Relief Valve Testing Page 10 of 10 Temperature Overpressure Protection for Light-Water Reactors,' Pursuant to 1 OCFR50.54(9," dated March 21, 1991.

4. J. F. Opeka letter B14198 to the U.S. Nuclear Regulatory Commission, "Haddam Neck Plant Millstone Nuclear Power Station, Unit Nos. 2 and 3 Generic Letter 90-06, Resolution of Generic lssue 70, 'Power-Operated Relief Valve and Block Valve Reliability,' and Generic lssue 94, 'Additional Low-Temperature Overpressure Protection for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f),11 dated August 20, 1992.

5. G. S. Vissing letter to J. F. Opeka, Staff Review Of Generic Letter 90-06, "Resolution Of Generic lssue 70, 'Power-Operated Relief Valve And Block Valve Reliability,' And Generic lssue 94, 'Additional Low-Temperature Overpressure Protection For Light-Water Reactors,' Pursuant To 10 CFR 50.54(f)," (TAC NOS.

M77361 AND M77431), dated November 6, 1992.

6. J. F. Opeka letter B14349 to the U. S. Nuclear Regulatory Commission, Millstone Nuclear Power Station, Unit No. 2 Resolution to Staff Review Comments on Generic Letter 90-06, "Resolution of Generic lssue 70, 'Power Operated Relief Valve and Block Valve Reliability,' and Generic lssue 94, 'Additional Low-Temperature Overpressure Protection for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f)," dated January 11, 1993.

7. J. F. Opeka letter B14559 to the U. S. Nuclear Regulatory Commission, Millstone Nuclear Power Station, Unit No. 2, Proposed Revision to Technical Specifications, Generic Letter 90-06, Dated April 25, 1994.

8. G. S. Vissing letter to J. F. Opeka, Issuance Of Amendment No. 185, (TAC No.

M89380), dated February 15, 1995.

Serial No. 07-0006 Docket No. 50-336 ATTACHMENT 2 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP2-008)

PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) TESTING MARKED-UP TECHNICAL SPECIFICATIONS PAGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS d

4.4.3.1 In addition to therequlreiments of Speclfication4.0.5, each PORV shall be demonstrated OPERABLE:

. a.

Once per 31 days by performance o f a CHANNEL FUNCTIONAL TEST, excluding valve operat Ion, and

b.

Once per 18 months by performance of a CHANNEL CALIBRATION.

1 4.4.3.2 Each block valve shall be demonstrated OPERABLE once per 92 days by operating the valve through one complete cycle of full travel.

This demonstration i s not required i f a PORV block valve i s closed and power removed t o meet Specfficatlon 3.4.3 b or c.

HILLSTONE - UNIT 2 0098

Serial No. 07-0006 Docket No. 50-336 Page 1 of I Insert A Millstone Power Station Unit 2 Page 314 4-3a Once per 18 months by operating the PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4.

Serial No. 07-0006 Docket No. 50-336 ATTACHMENT 3 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS (LBDCR 05-MP2-008)

PRESSURIZER POWER OPERATED RELIEF VALVES (PORVs) TESTING MARKED-UP BASES PAGE DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

314.4 REACTOR COOLANT SYSTEM BASES stuck open PORV at a time that the block valve is inoperable. This may be accolnplished by various methods. These methods include, but are not limited to, placing the NORMAL/ISOLATE switch at the associated Bottle Up Panel in the "ISOLATE" position or pulling the control power fuses for the associated PORV control circuit.

Although the block valve may be designated inoperable, it may be able to be ma~~ually opened and ctosed and in this manner can be used to perform its function. Block valve inoperability may be due to seat leakage, instrumentation problems, or other causes that do not prevent manual use and do not create a possibility for a small break LOCA. This condition is only intended to permit operation of the plant for a limited period of time. The block valve should normally be available to allow PORV operation for automatic mitigation of overpressure events. The block valves must be returned to OPERABLE status prior to entering MODE 3 after a refueling outage.

If more than one PORV is inoperable and not capable of being manually cycled, it is necessary to either restore at least one valve within the completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or isolate the flow path by closing and removing the power to the associated block valve.and cooldown the RCS to MODE 4.

C 3/4.4.4 PRESSURIZER An OPERABLE pressurizer provides pressure control for the reactor coolant system during operations with both forced reactor.coo1ant flow and with natural circulation flow. The minimum water level in the pressurizer. assures the pressurizer heaters, which are required to achieve and maintain pressure control, remain covered with water to prevent failure, which occurs if the heaters are energized uncovered. The maximum water level in the pressurizer ensures that this parameter is maintained within the envelope of operation assumed in the safety analysis. The maximum water level also ensures that the RCS is not a hydraulically solid system and that a steam bubble will be provided to accommodate pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish and

- maintain natural circulation.

The reqGrement for two groups of pressurizer heaters, each having a capacity of 130 kW, is met by verifying the capacity of the preqsurizer proportional heater groups 1 and 2. Since the pressurizer proportional heater groups 1 and 2 are supplied from the emergency 480V electrical buses, there is reasonable assurance that these heaters can be energized during a loss of offsite power to maintain natural circulation ~ ~. W O T STANDBY.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a n~odification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is MILLSTONE - UNIT 2 B 314 4-2a Amendment No. 2?,37,52,66,97,

Serial No. 07-0006 Docket No. 50-336 Page 1 of 1 Insert B Millstone Power Station Unit 2 Page B314 4-2a For Information Only SURVEILLANCE REQUIREMENT 4.4.3.1.c requires operating each PORV through one complete cycle of full travel at conditions representative of MODES 3 or 4. This is normally performed in MODE 3 or 4 as the unit is descending in power to commence a refueling outage. This test will normally be a static test, whereby a PORV will be exposed to MODE 3 or 4 temperatures, the block valve closed, and the PORV tested to verify it strokes through one complete cycle of full travel. PORV cycling demonstrates its function. The Frequency of 18 months is based on a typical refueling cycle and industry accepted practice. SURVEILLANCE REQUIREMENT 4.4.3.1.c is consistent with the NRC staff position outlined in Generic Letter 90-06, which requires that the 18-month PORV stroke test be performed at conditions representative of MODE 3 or 4.

Testing in the manner described is also consistent with the guidance in NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants," Section 4.2.10, that describes the PORVs function during reactor startup and shutdown to protect the reactor vessel and coolant system from low-temperature overpressurization conditions, and indicates they should be exercised before system conditions warrant vessel protection. If post maintenance retest is warranted, the affected valve(s) will be stroked under ambient conditions while in Mode 5, 6, or defueled. The actual stroke time in the open and close direction will be measured, recorded and compared to the test results obtained during pre-installation testing to assess acceptability of the affected valve(s).