ML17251A474

Rev 0 to Safety Analysis,Ginna Station PORV Block Valves.

ML17251A474
Person / Time
Site:	Ginna
Issue date:	04/01/1987
From:	Pham C ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17251A473	List: ML17251A472 ML17251A474 ML17251A477 ML17251A478 ML17251A479
References
EWR-3755-01, EWR-3755-1, EWR-3755-R, EWR-3755-R00, NUDOCS 8905160061
Download: ML17251A474 (16)

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Safety Analysis Ginna Station PORV Block Valves Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 EWR 3755 Revision 0 January 22, 1986 Prepared by: C. H.

Mech 'l h~.~r Engineer ra OATK Reviewed by: 8'ATE clear Engineer Approved by:

OATE Manager, Nuclear Engineering Page i S A13H 890516006i 890503 PDR,..ADOCK 05000244 P PDC

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1.0 Sco e of Anal sis This analysis covers the replacement of PORV block valves 515 and 516.

1.2 Protection against overpressurization in the Reactor Coolant System at Ginna Station is provided by two spring-loaded safety valves and two power-operated relief valves (PORV). The lines leading to eachblock PORV also contain remotely-actuated, motor-operated valves to be used if the PORV sticks open. Since each block valve serves as a backup means of limiting reactor coolant loss in the event that a transient challanges the POPV's and a PORV subsequently fails open, leakage limits have been established for the block valves which are consistent with the makeup capacity provided by a single charging pump. Leakage in excess of these limits could require a plant shutdown. To maintain block valve seat leakage within these limits, been necessary to periodically disassemble the it has valves and remachine the seating surfaces. The block valves (V515 and V516) seat rings are now approaching the maximum allowable limits for remachining.

1.3 The purpose of the modification is to replace the two block valves (515 and 516) with new ASME N-stamped, motor-operated block valves which are seismically and environmentally qualified. The valves are located inside containment on top of the Reactor Coolant System pressurizer.

2.0 Referenced Documents 2.1 Design Criteria, EWR 3755, Revision 0, January 22, 1986.

2.2 Ginna Station Updated Final Safety Analysis Report, (UFSAR).

2.2.1 Chapter 3, "Design of Structures, Components, Equipment and Systems".

2.2.2 Chapter 5, "Reactor Coolant System and Connected Systems".

2.3 Ginna Technical Specifications 2.3.1 Section 3.1, "Limiting Conditions for Operation - Reactor Coolant System".

2.3.2 Section 4.1, "Surveillance Requirements - Operational Safety Review".

Safety Analysis Revision Page 1 EWR 3755 Date 1 22 86

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2.3.3 Section 4.16, "Surveillance Requirements - Overpressure Protection System".

2.3.4 Section 5.3.2, "Design Features - Reactor Coolant System".

2.4 Code of Federal Regulations - Title 10, Part 50 (10CFR50).

2.4.1 Appendix A, "General Design Criteria for Nuclear Power Plants".

2.5 Ginna Station Quality Assurance Manual, Appendix A, "Quality and Safety Related Systems Listing and Diagrams,"

Rev. 5, May 15, 1986.

2.6 USNRC, Regulatory Guide 1.70 "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", Rev. 3,,November 1978.

3.0 Safet Anal sis 3.1 A review has been made of all events analyzed in the Ginna Station UFSAR and the events requiring analysis by NRC Regulatory Guide 1.70. The events related to this modification are:

1) Increase in Heat Removal by Secondary System.

2) Decrease in Heat Removal by Secondary System.

3) Decrease in Reactor Coolant System Flow Rate.

4) Reactivity and Power Distribution Anomalies.

5) Increase in Reactor Coolant Inventory.

6) Decrease in Reactor Coolant Inventory.

7) Internal and External Events such as Major and Minor Fires, Floods, Storms, or Earthquakes.

3.2 Increase in Heat Removal by Secondary System 3.2.1 The following accidents, applicable to this event, were analyzed:

1) Decrease in Feedwater Temperature

2) Increase in Feedwater Flow Safety Analysis Revision 0 Page 2 EWR 3755 Date 1 22 86

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3) Excessive Load Increase Incident Inadvertant Opening of a Steam Generator Relief/

Safety Valve

5) Spectrum of Steam System Piping Failures Inside and Outside of Containment 3.2.2 This modification will not, degrade the design, capability or per formance o f the existing pressurizer relic f system and, therefore, the consequences of this event will not be increased by the modification 3.3 Decxease in Heat Removal by Secondary System 3.3.1 The following accidents, applicable to this event, were analyzed:

1): Steam Pressure Regulator Malfunction or Failure that Results in Decxeasing Steam Flow

2) loss of External Electrical Load

3) Turbine Trip

4) Loss of Condenser Vacuum

5) Loss of Offsite Alternating Current Power to the Station Auxiliaries

6) Loss of Normal Feedwater Flow

7) Feedwater System Pipe Breaks 3.3.2 This modification will not degrade the design, capability or performance of the existing pressurizer relief system and, therefore, the consequences of this event will not be increased by the modification.

3.4 Decrease in Reactor Coolant System Flow Rate 3.4.1 The following accidents, applicable to this event, were analyzed:

1) Flow Coastdown Accidents

2) Locked Rotor Accidents 3.4.2 This modification will not. degrade the design, capability or performance of the existing pressurizer relief system and, therefore, the consequences of this event will not be increased by the modification.

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3.5 Reactivity and Power Distribution Anomalies 3.5.1 The following accidents, applicable to this event, were analyzed:

1) Uncontrolled Rod Cluster Control Assembly Withdrawal from a Subcritical Condition

2) Uncontrolled Rod Cluster Control Assembly Withdrawal at Power

3) Startup of an Inactive Reactor Coolant Pump

4) .Chemical and Volume Control System Malfunction

5) Rupture of a Control Rod Drive Mechanism Housing

6) Rod Cluster Control Assembly Drop 3.5.2 This modification will not degrade the design, capability or performance of the existing pressurizer relief system and, therefore, the consequences of this event will not be increased by the modification.

3.6 Increase in Reactor Coolant Inventory 3.6.1 This modification will not degrade the design, capability or performance of the existing pressurizer relief system and, therefore, the consequences of this event will not be increased by the modification.

3.7 Decrease in Reactor Coolant Inventory 3.7.1 The following accidents, applicable to this event, were analyzed:

1) Inadvertant Opening of a Pressurizer Safety or Relief Valve

2) Primary System Pipe Ruptures 3.7.2 This modification will not degrade the design, capability or performance of the existing pressurizer relief system and, therefore,-the consequences of this event will not be increased by the modification.

3.8 Internal and External Events such as Major and Minor Fires, Floods, Storms, or Earthquakes.

3.8.1 Fires Safety Analysis Revision Page 4 EWR 3755 Date 1 22 86

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3 ~ 8 ~ 1.1 A review of this modification per Engineering Procedure QE-326 will ensure compliance with 10CFR50, Appendix R.

3 ~ 8 ~ 2 Flood and Storms 3 ' '.1 The any modification neither affects, nor is affected by, flood or storm previously evaluated.

3.8.3 Earthquake 3.8 '.1 This modification is Seismic Category 1.

3.8.4 Environmental Conditions The PORV block valves and operator shall be qualified for operation within the environmental condition specified in the Design Criteria which envelope those specified in section 3.11.3 of the Ginna USFAR.

3.9 Ba ed on the above analysis, it can be concluded that the margins of safety during normal operations and transient conditions anticipated during the life of the plant have not been reduced. The adequacy of structures, systems, and components provided for the prevention of accidents and for the mitigation of the consequences of accidents have not been affected.

4.0 Prel imina ft Sa e Eva lug tion 4.1 The probability of'occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report will not be increased by the proposed modification.

4.2 The possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis will not be created by the proposed modification.

4.3 The margin of safety as defined in the basis for any Technical Specification will not be reduced by the proposed modification.

4.4 The proposed modification does not require a Technical Specification change.

4.5 The proposed modification does not involve an unreviewed safety question.

Safety Analysis Revision Page 5 ENR 3755 Date 1/22/86

0 ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION PORV BLOCK VALVE REPLACEMENT PROGRAM ATTACHMENT C. 3 A copy of the applicable portions of the EPRI Report, "EPRI PWR Safety and Relief Valve Test Program PORV Block Valve Information Package", dated May 31, 1982;