ML20065F573

Application for Amend to License NPF-3,adding TS 3/4.4.12 Limiting Condition for Operation,Surveillance Requirements & Bases Section for Pilot Operated Relief Valve & Block Valve

ML20065F573
Person / Time
Site:	Davis Besse
Issue date:	03/30/1994
From:	Shelton D CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20065F580	List: ML20065F573 ML20065F584
References
2211, NUDOCS 9404110421
Download: ML20065F573 (14)

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u CENTERDOR W ENERGY 6200 00k Tree Boulevard Matl Adoress: Donald C. Shelton Independence OH P.O. Box 94661 Senior Vice President 216 447-3153 Cleveland, OH 44101-4661 Nuclear Fox 216-447-3123 Docket' Number 50-346 License Number NPF-3 Serial Number 2211 March 30.1994 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-

Subject:

Proposed Modification to the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A Technical Specifications to Add a New Section 3/4.4.12, Pilot operated Relief Valve (PORV) and Block Valve Gentlemen Enclosed is an application for an amendment to the DBNPS Unit Number 1 operating License Number NPF-3, Appendix A, Technical Specifications to reflect the changes attached. This application proposes adding a new Technical Specification 3/4.4.12 Limiting Condition for Operation, Surveillance Requirements and Bases Section for the Pilot. Operated Relief Valve (PORV) and its block valve. The changes proposed by this amendment application are in general conformance with Generic Letter 90-06 (GL 90-06) recommendations issued by the NRC on June 25, 1990.

Deviations from the guidelines recommended in Enclosure A to GL 90-06 were discussed with the NRC at a meeting between'the NRC and. Toledo Edison on July 29, 1993 and are covered in detail in the attached.

Safety Assessment and Significant Hazards Consideration.

This License Amendment applicaElon supersedes the Toledo Edison request for changes to Technical Specification Bases-3/4.4.11. High Point Vents, submitted to the NRC by Toledo Edison letter Serial Number 2046 dated June 26, 1992. As stated in Toledo Edison letter Serial Number 2191 dated November 30, 1993, Toledo Edison withdraws the proposed change to Technical Specification Bases 3/4.4.11, High Point Vents.

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Operating Companies:

Clevelond Electric tttuminchng 9404110421 940330 I' Toledo Edison PDR ADOCK 05000346- #

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Docket Number 50-346 License Number NPF-3 Seri'al Number 2211 Page 2 Should further information be required, please contact Mr. William T. O'Connor, Manager - Regulatory Affairs, at (419) 249-2366.

Ver truly yours,

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's DCS FWK Enclosure cc: J. B. Martin, Regional Administrator, NRC Region III S. Stasek, NRC Region III, DB-1 Senior Resident Inspector G. West Jr. , DB-1 NRC/NRR Project Manager J. R. Williams, Chief of Staff, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board l

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Docket Number 50-346 License Number NPF-3 Serial Number 2211 Enclosure Page 1 APPLICATION FOR AMENDMENT TO 1 l

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FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3. Also included is the Safety Assessment and Significant Hazards Consideration.

The proposed changes (submitted under cover letter Serial Number 2211)

Concernt Appendix A, Technical Specification Section 3/4.4.12, Pilot Operated Relief. Valve and Block Valve.

Appendix A. Technical Specification Bases 3/4.4.12, Pilot Operated Relief Valve and Block Valve.

Fort D.C. Shelton, Senior Vice President - Nuclear Bys f crb Og J. K. Nood, Plant Manager Sworn to and subscribed before me this 30th day of March,1994, 11/l/Ja) lA$

Notary Pul5fic. State of Ohio

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EVELYN LOMSS NOT;hv Tts.10, STATE OF Chic HyOsa:nEr$cs#f2. W l

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. Docket Number 50-346 License Number NPF-3 Serial Number 2211 Enclpsure' Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Operating License Number NPF-3, Appendix A. Technical Specifications.

The change involves the addition of a new Technical Specification Section 3/4.4.12, Pilot Operated Relief Valve (PORV) and Block Valve, and its associated Bases.

A. Time Required to Implement: This change is to be implemented within 90 days after the NRC issuance of the License Amendment.

B. Reason for Change (License Amendment Request Number 93-0007).

To comply with the recommendations of Generic Letter 90-06 and resolve Generic Issue 70, " Power Operated Relief Valve and Block Valve Reliability" for the Davis-Besse Nuclear Power Station.

C. Safety Assessment and Significant Hazards Considerations See Attachment.

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Docket Number 50-346 License Number NPF-3 Serial Number 2211 Attachment Page 1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION i FOR LICENSE AMENDMENT REQUEST NUMBER 93-07 TITLE:

Proposed Revision to the Davis-Besse Nuclear Power Station (DBNPS)

Operating License NPF-3 Appendix A Technical Specifications to Add a New Technical Specification 3/4.4.12, Pilot Operated Relief Valve (PORV) and Block Valve.

DESCRIPTION:

The Nuclear Regulatory Commission (NRC) issued Generic Letter 90-06 (GL 90-06) Resolution of Generic Issue (GI) 70, " Power-Operated Relief Valve and Block Valve Reliability," and Generic Issue 94 " Additional Low-Temperature Overpressure Protection for Light-Water Reactors." The NRC's technical findings and the regulatory analysis related to GI-70 were discussed in NUREG-1316, " Technical Findings and Regulatory Analysis Related to Generic Issue 70 - Evaluation of Power-Operated Relief Valve and Block Valve Reliability in PWR Nuclear Power Plants."

In GL 90-06, the NRC advised that it had completed an effort to evaluate the role of PORVs in performing certain safety-related functions. The 2

NRC determined that over a period of time, the role of PORVs are now relied upon by various plants to perform one, or more, of the following safety-related functions:

1. Hitigation of a design-basis steam generator tube rupture (SGTR) accident,

2. Low-temperature overpressure protection of the reactor vessel during startup and shutdown, or

3. Plant cooldown in compliance with Branch Technical Position Reactor Systems Branch (RSB) 5-1 to Standard Review Plan (SRP) 5.4.7, " Residual Heat Removal (RHR) System."

4 The NRC recommended that Technical Specifications be adopted to increase the reliability and availability of the PORV and block valve during performance of one or more of the above safety related functions.

Enclosure A to GL 90-06 provided Technical Specification guidance for B&W type plants regarding Generic Issue 70, " Power-Operated Relief Valve and Block Valve Reliability." However, as noted by the NRC in GL 90-06, Generic Issue 94 " Low-Temperature Overpressure Protection" is not applicable to B&W-designed plants like the DBNPS and therefore, need not be addressed.

Docket Number 50-346 License Number NPF-3 Serial Number 2211 Attachment Page 2 ]

l This license amendment request is to modify the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3 Appendix A Technical l Specifications by adding a new Technical Specification Limiting Condition for Operation, Surveillance Requirements and Bases Section for the PORV and block valve as recommended by GL 90-06 Attachment A, and resolve Generic issue 70, " Power-Operated Relief Valve and Block Valve 1 Reliability" for the DBNPS. The content of these proposed Technical'

! Specifications were discussed with the NRC at a meeting between the NRC

( and Toledo Edison on July 29, 1993.

! The-proposed addition to the DBNPS Technical Specifications will add l Limiting Condition for Operation (LCO) 3.4.12 that states, "The Pilot

! Operated Relief Valve (PORV) and its associated block valve shall be l OPERABLE." The LCO is applicable in MODES 1 (Power Operation),

l 2 (Startup), and 3 (Hot Standby).

There are four ACTION requirements associated with the proposed LCO 3.4.12. They ares

a. With the PORV inoperable because of seat leakage, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

! restore the PORV to OPERABLE status or close the PORV block valve

! with power maintained to the block valve; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION a. ensures that the integrity of the associated reactor coolant boundary will be preserved and that failure of the PORV will not become the initiator of a Reactor Coolant System (RCS) Loss of Coolant Accident (LOCA).

b. With the PORV inoperable due to causes other than seat leakage:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the PORV block valve.

2. If the PORV block valve is closed to comply with ACTION b.1, restore the PORV to OPERABLE status within the following 7 days and reopen the block valve, or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With the block valve inoperable, within 7 days restore the block l valve to OPERABLE status; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTIONS b. and c. limit the unavailability of the PORV during MODES 1, 2, and 3.

d. The provisions of Specification 3.0.4 are not applicable.

ACTION d. allows entry into the applicable modes while relying upon the provisions contained in the ACTION statements.

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. .- Docket Number-50-346 License Number NPF-3 Serial Number 2211

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Attachment Page 3 Two new Surveillance Requirements have been added:

4.4.12.1 In addition to the requirements of Specification 4.0.5, the PORV shall be demonstrated OPERABLE at least once per 18 months by operating the PORV through one complete cycle of full travel during MODES 3 or 4.

4.4.12.2 The PORV Block Valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of Specification 3.4.12 ACTION a. or b.

This license amendment request follows the NRC's Technical Specification recommendations of Enclosure A to GL 90-06 with the following exceptions:

1. ACTION a.

The action time allowance to restore the PORV or close the associated block valve is increased from I hour to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to be consistent with DBNPS RCS Operational Leakage LCO 3.4.6.2.

Action b. The NRC's Improved Standard Technical Specifications (NUREG-1430) LCO 3.4.13 also has a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allowance for operational leakage.

2. ACTION b.

The action does not require removal of power from the block valve once it has been closed. With the block valve closed the associated Reactor Coolant System integrity is assured. The closed block valve provides protection from a loss of coolant through the PORV flowpath. Maintaining the block valve energized will enhance the availability of the PORV for feed and bleed cooling under those circumstances where the PORV though inoperable, is still functionally capable of passing the required flow needed for the beyond - design basis event feed and bleed cooling.

The action time allowance is increased from 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to 7 days once the block valve is closed. This 7 day allowance will provide time for minor repairs that allow returning the PORV to operable status without requiring shutdown. It takes into consideration the DBNPS's diverse Reactor Coolant System pressure control available for recovery from a Steam Generator Tube Rupture (SGTR) event. This diversity includes the use of the pressurizer spray as the primary means of RCS depressurization, or the use of the pressurizer vent line (valves RC239A and RC200) in the event of a loss of offsite power.

The NRC's letter of October 14, 1992 to Toledo Edison (Toledo Edison Log No 3850), stated:

Docket Number 50-346

. License Number NPF-3 Serial Number 2211 Attachment Page 4 As discussed in NUREG-1316, most of the safety enhancement for the proposed backfit is derived from the increase in feed and bleed _ capability. Therefore, for those facilities that can accomplish the_three safety functions identified in the. generic letter without relying on PORVs for the feed and bleed function, the allowed outage time (A0T) for an inoperable PORV (for reasons otherlthan excessive leakage) may be increased from the recommended ,

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days.

For the DBNPS, feed and bleed cooling would only be' required in a beyond-design basis event involving the loss of~

primary-to-secondary' heat transfer (e.g., loss of both main feedwater pumps, loss of both turbine' driven auxiliary.

feedwater' pumps, and loss of-the motor driven feedwater pump). ,

The previously existing feed and bleed cooling capability ,

required the use of the PORV, RCS pressurizer ASME Code safety i valves (PSVs), and both makeup pumps. However, upgrades to the Makeup'(MU) System, as described'in Toledo. Edison'se l'tter to the NRC dated September 18, 1990 (Serial Number'1836), have provided increased flow capability, train independence, .

reduction of common mode failure probability, and functionality following a seismic event and a loss of offsite power. As a result of these upgrades, feed and bleed cooling will not be-lost upon a failure of either a MU pump or the PORV.

Analytical results indicate'that successful feed and bleed cooling will be attained with the following minimum equipment i combinations:

t Two makeup pumps and the RCS pressurizer. code safety .

valves, or "

One makeup pump operating in piggyback with a Low Pressure Injection Pump, RCS pressurizer code safety valves, and the PORV.

l For the DBNPS, these upgrades, therefore, have allowed for the loss of the PORV flowpath without the loss of feed and bleed cooling capability. The loss of the PORV flowpath does not affect the capability.to properly cool the core as long as two makeup pumps and the PSVs are available. Technical Specification operability requirements. exist for the two MU pumps (TS 3.1.2.4) and the PSVs (TS 3.4.3).

3. ACTION c.

The action time allowance to restore the block valve to operable status is increased from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 7 days. This allows time to perform minor repairs. Extending the action time to 7 days is acceptable at the DBNPS due to the diversity; of options available for RCS depressurization (as discussed above) and the unlike11 hood of the PORV being actuated and then becoming stuck in an open position during this allowable outage time.

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Docket Number 50-346 License Number NPF-3 Serial Number 2211 Attachment Page 5 At the DBNPS the PORV does not have a " manual-closed" position and must remain in automatic (AUT0). However, the PORV setpoint has been raised above the RCS High Pressure trip setpoint to 2435 psig or greater in accordance with Technical Specification 3/4.4.3, RCS-Safety Valves and PORV. The RCS High Pressure trip is set at 2355 psig in accordance with Technical Specification 2.1.1, Safety Limits-Reactor Core.

This ensures that during any anticipated high pressure transient, the reactor will trip to prevent the pressure overshoot from reaching the PORV setpoint. Technical Specification Bases Section 3/4.4.3 " Safety Valves" explains the following: "The pressurizer pilot operated relief valve should be set such that it will open before the code safety valves are opened. However, it should not open on any anticipated transients. BAW-1890, September 1985 identified that the turbine trip from full power would cause-the-largest overpressure transient. This report demonstrated that with a RPS high pressure trip setpoint of 2355 psig the resulting overshoot in RCS pressure would be limited to 50 psi." Adding the instrument string error provides a minimum PORV setpoint of 2435 psig. The field setpoint is 2450 psig to ensure the PORV will not open before 2435 psig. Therefore, the PORV can be expected to maintain the reactor coolant boundary while the switch position is in AUTO.

In addition, the safety-grade Anticipatory Reactor Trip System trip is initiated by a turbine trip (above 45 percent of rated thermal power) or trip of both main feedwater pump turbines.

This trip will operate in advance of the RCS high pressure trip to reduce the peak RCS pressure and thus reduce challenges to the PORV.

4. NRC recommended Surveillance Requirement 4.4.4.1 regarding channel calibration is already covered by existing DBNPS Surveillance 4.4.3 (RCS- Safety Valves and PORV - Operating; applicable in Modes 1, 2 and 3) on an 18 month frequency and, therefore, is not repeated here.

5. NRC recommended Surveillance Requirement 4.4.4.3 regarding the testing of the transfer to emergency power is not included since Class 1E essential power is already the normal power supply to the PORV and block valve at the DBNPS.

A new Bases 3/4.4.12, Pilot Operated Relief Valve and Block Valve, has been added to explain the purpose of the new Technical Specification. In addition, the Technical Specifications Index pages have been proposed for revision to reflect the new Technical Specification and its Bases.

. Docket Number 50-346 License Number NPF-3 Serial Number 2211 Attachment l Page 6 l SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED: l l

These proposed Technical Specification changes affect the pilot-operated relief valve (PORV) and its associated block valve. The DBNPS Babcock and Wilcox-designed Nuclear Steam Supply System contains one PORV and one block valve. The PORV (RC2A) is a Crosby, solenoid actuated valve. The block valve (RC11) is a Velan gate valve with a Limitorque valve operator.

SAFETY FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES:

The safety function of the PORV is to provide RCS pressure relief and act as a reactor coolant pressure boundary. The function of the block valve is to provide a reactor coolant pressure boundary for an inoperable PORV.

The NRC advised in GL 90-06 that over a period of time, the role of PORVs had changed such that PORVs are now relied upon by various plants to perform one, or more, of the following safety-related functions:

1. Mitigation of a design - basis steam generator tube rupture accident,

2. Low-temperature overpressure protection of the reactor vessel during startup and shutdown, or

3. Plant cooldown in compliance with Branch Technical Position RSB 5-1 to SRP 5.4.7, " Residual leat Removal (RHR) System."

The following discusses each of these three items:

1. Hitigation of a design-basis steam generator tube rupture (SGTR) accident at the DBNPS can utilize the PORV during the recovery sequence, however, its use is not the primary method of recovery.

In accordance with emergency procedure DB-OP-02000 "RPS, SFAS, SFRCS Trip or Steam Generator Tube Rupture," control of the event is accomplished by steaming the steam generators (SGs) through the turbine bypass valves (TBVs) or through the atmospheric vent valves (AVVs) if the condenser is not available. The TBVs and AVVs are controlled to assure that SG pressure is kept below the main steam safety valve (MSSV) lift setpoint. The RCS is depressurized by turning off the pressurizer heaters and using pressurizer spray.

Pressurizer spray can be credited since the analysis does not assume loss-of-offsite power in USAR Section 15.4.2 " Steam Generator Tube Rupture." As a result, the RCPs are available for pressurizer spray for pressure control. However, if the pressurizer spray is not available, then the use of the pressurizer vent line (valves RC239A and RC200) is directed by the procedure as the primary means of RCS pressure reduction. The use of the pressurizer vent line (RC239A and RC200) for pressure control during recovery from a SGTR was chosen because a restricting orifice in this vent line limits flow to provide a more controlled pressure reduction with less risk of unintentionally rupturing the pressurizer quench tank rupture disk.

Both the PORV and the pressurizer vent line discharge to the pressurizer quench tank, but the PORV has the capability of a mass

-' Docket Number 50-340 License Number NPF-3 Serial Number 2211 Attachment ,

Page 7 flow rate sufficient to rupture the pressurizer quench tank rupture disk. No credit is taken in the DBNPS Updated Safety Analysis Report for the usage of the PORV and block valve path or pressurizer vent line for accomplishing RCS depressurization during a SGTR, and the calculated resulting doses are well within the 10CFR Part 100 guidelines.

2. Low-temperature overpressure protection (LTOP) is not an issue for B&W designed plants like the DBNPS as noted in GL 90-06 and NUREG-1316 therefore, LCOs for the PORV and block valve are not needed for LTOP considerations at the DBNPS.

3. Regarding safety grade cooldown, the DBNPS received its Operating License in April 1977 and, therefore, is not licensed to Branch-Technical Position RSB 5-1 in Standard Review Plan 5.4.7, Residual Heat Removal System. The DBNPS is not required to meet the branch technical position regarding safety grade cooldown and, in fact, the DBNPS is designed to remain in hot standby during natural circulation operation. DBNPS procedure DB-OP-06903, " Plant Shutdown and Cooldown", directs maintaining the plant in hot; standby during natural circulation. If plant conditions force a natural circulation cooldown, procedure DB-OP-06903, " Plant Shutdown and Cooldown", restricts the cooldown rate to 100F per hour in order to prevent forming a steam space in the reactor vessel head. The use of the PORV flowpath (RC2A and RCll) for pressure control during a natural circulation cooldown (i.e. loss-of-offsite power and loss of Reactor Coolant Pumps) is not desired at the DBNPS and is not specified in the procedure. DB-0P-06903, Section 7.0, "Cooldown on Natural Circulation" directs the use of the pressurizer vent line (RC239A and RC200). The use of the pressurizer vent line for depressurization during natural circulation cooldowns is desired since the pressurizer vent line's depressurization rate is less than that of the PORV flowpath and, therefore, more controllable with less risk of unintentionally rupturing the pressurizer quench tank rupture disk.

At the DBNPS the PORV can be used to perform other functions, they are:

1. Beyond-Design Basis Event " Feed and Bleed" Core Cooling.

For the DBNPS, feed and bleed cooling would only be required in a beyond-design basis event involving the loss of primary-to-secondary heat transfer (e.g., loss of both main feedwater pumps, loss of both  ;

turbine driven auxiliary feedwater pumps, and loss of the motor j driven feedwater pamp). The previously existing feed and bleed cooling capability required the use of the PORV, RCS pressurizer .

ASME Code safety valves (PSVs), and both makeup pumps. However, upgrades to the Makeup (MU) System, as described in Toledo Edison's letter to the NRC dated September 18, 1990 (Serial Number 1836),

have provided increased flow capability, train independence, reduction of common mode failure probability, and functionality following a seismic event and a loss of offsite power. As a result of these upgrades, feed and bleed cooling will not be lost upon a

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failure of either a MU. pump or the PORV. Analytical results )

l indicated that successful feed and bleed cooling will be attained l with the following minimum equipment combinations: ,

l Two makeup pumps and the RCS pressurizer code safetyLvalves, or One makeup pump operating in. piggyback with a Low Pressure Injection Pump,.RCS pressurizer code safety valves,~and the-PORV.

l I For the DBNPS, these upgrades, therefore, have all'wed'for o the loss l of the PORV flowpath without the. loss of feed.and bleed cooling

, capability. ;The loss of.the PORV.flowpath does not affect the capability.to properly' cool the core as.long as two makeup pumps.and the PSVs are available.

2. Removal of non-condensible gases from the pressurizer following an l . accident by using the FORV as one redundant method.

USAR Section 6.3.3.1.4, Discussion of Noncondensible Gases.. explains that small amounts of noncondensible gases can be released into the; primary system'during a small break accident. Although the probability for such an occurrence is believed to be small, the pressurizer vent path (through either valves RC11 and RC2A, or valves RC239A and RC200) will provide a means of venting these non-condensible gases and ensures.the availability of an alternate means of depressurizing the RCS in the event of a steam generator tube rupture or natural circulation cooldown.

3. Prevention of challenges to the pressurizer ASME: Code safety valves.

4 The PORV setpoint is set to open prior to' reaching the setpoint of the pressurizer ASME Code safety valves.' Thus,'the PORV will function to relieve pressure and prevent the actuation of-the Code safety valves.

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The PORV's block valve (RC11) is provided to isolate the PORV flowpath should the PORV be inoperable in the open position or to-isolate the PORV-with a:1eaking disk. This prevents uncontrolled'depressurization and.

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excessive. leakage of reactor coolant.

EFFECTS ON SAFETY:

The' proposed changes to the DBNPS Technical Specifications to add a new Technical Specification Limiting condition for Operation (LCO),

Surveillance Requirements.and Bases'Section follows the NRC position with.

regards to. adding a new Technical Specification to address the NRC's concerns in Generic Letter 90-06 and Generic Issue 70. This new Technical Specification provides specific surveillances to ensure PORV and block valve operability. It also provides specific actions that address an inoperable PORV due to. seat leakage or other reasons, and an

i Docket Number 50-346 l bicense Number NPF-3 Serial Number 2211 Attachment Page 9 inoperable block valve. Allowed outage times have been proposed to I minimize unnecessarily putting the plant through cooldowns and heatups.

j These are based on discussions with the NRC staff and are derived from i engineering and operational judgment.

New Surveillance Requirement 4.4.12.1, which requires the operation of I the PORV through one complete cycle of full travel during Mode 3 or 4 at least once per 18 months +25% is already being accomplished in accordance l

with procedure DB-SP-03363, Pressurizer Power Operated Relief Valve Cycle Test. The PORV is cycled through full travel in Mode 3 during the cooldown to or the subsequent heatup from Mode 5 if the PORV has not been cycled within the previous 3 months. Therefore, this new surveillance l Requirement does not represent an actual increase in testing at the DBNPS.

New Surveillance Requirement 4.4.12.2, which requires that the PORV's block valve be operated through one complete cycle of full travel once per 92 days, is already being performed in accordance with the DBNPS Inservice Testing Program. Therefore, this new Surveillance Requirement does not represent an actual increase in testing at the DBNPS.

This application also includes Bases and Index changes. These changes are associated administrative changes and do not affect safety.

Accordingly, it is concluded that this proposed change.has no adverse effect on plant safety.

SIGNIFICANT HAZARDS CONSIDERATION:

The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment to an Operating License for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would: (1) Not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) Not )

create the possibility of a new or different kind of accident from any l accident previously evaluated; or (3) Not involve a significant reduction  ;

in a margin of safety. Toledo Edison has reviewed the proposed changes and determined that a significant hazards consideration does not exist l because operation of the Davis-Besse Nuclear Power Station, Unit Number 1, in accordance with these changes would: I la. Not involve a significant increase in the probability of an accident I previously evaluated because no change is being made to any accident initiator. Automatic actuation of the PORV is not assumed to l mitigate the consequences of a design basis accident as described in Chapter 15 of the USAR. The proposed changes will continue to ensure the PORV and block valves are available to perform their functions when required to do so. Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

Docket Number 50-346 License Number NPF-3 Serial Number 2211 Attachment Page 10 lb. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed changes do not invalidate accident conditions or assumptions used in evaluating the radiological consequences of an accident.

2a. Not create the possibility of a new kind of accident from any accident previously evaluated because the proposed changes do not delete any function previously provided by the PORV nor has the possibility of inadvertent opening been increased. No new types of s failures or accident initiators are introduced by the proposed changes.

2b. Not create the possibility of a different kind of accident from any accident previously evaluated because no new failure modes have been defined for any plant system or component important to safety, nor has any new limiting single failure been identified as a result of the proposed changes. No different accident initiators or failure mechanisms are introduced by the proposed changes.

3. Not involve a significant reduction in a margin of safety because the proposed changes continue to ensure the availability of the PORV and block valve when called upon to perform their function and will

not impact any safety analysis sssumptions.

CONCLUSIONS:

On the basis of the above, Toledo Edison has determined that the License Amendment Request does not involve a significant hazards consideration.

As this License Amendment Request concerns a proposed change to the Technical Specifications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.

ATTACHMENT:

Attached are the proposed marked-up changes to the Operating License.