ML20077E511

Application for Amend to License NPF-3,removing Tech Specs Decay Heat Isolation Valve & Pressurizer Heater Interlock Requirements in Modes 4 & 5.Proposed Changes Involve Tech Specs 3/4.3.2

ML20077E511
Person / Time
Site:	Davis Besse
Issue date:	06/04/1991
From:	Shelton D CENTERIOR ENERGY
To:
Shared Package
ML20077E505	List: ML20077E502 ML20077E511 ML20077E517
References
NUDOCS 9106100425
Download: ML20077E511 (7)

Text

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.' Docket Number 5D-346 License Number NPF-3 Serial Number 1018 Enclosure Page 1 APPLICATION FOR AMENDMEFT TO FACILITY OPERATING 1.ICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 Attached is a requested change to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Numbet NPF-3 Appendix A, Technical Specifications. Also included is the Safety Assessment and Significant Hazards Consideration.

The proposed change (submitted under cover letter Ser'al Number 1918) concerns:

Technical Specification 3/4.3.2, Safety Featutes Actuation System Ins t r umentation ny, /~? b Q D. C'. Sheltor, Vice President -

Nuclear, Davis-Besse Svoin and sul cribe* before me this 4th day of June 1991.

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Notary Publi , State of Ohio EVELYNL DRESS N'yTARY PUE STATEOF OH10 My w= E2;*ss My 3.tifA 9106100425 16604 FDR fiUOCr ' ' 05000 34 6 T' PDR

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. Docket Number 50-346 License Numbet NPF-3 1 Serial Number 1918 tnclosute 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 l License Number NPF-3, Appendix A, Technical Specifications. Technical i Specif1 cation 3/4.3.2.

• A. Time required to implement: This change is to be implemented vithin 45 days after NRC issuance of the License Amendment by the NRC.

B. Reason for change (License Amendment Request Number 91-0006): The requested changes remove the operability requirements for decay heat l isolation valve and pressurizer heater interlocks in Modes 4 and 5.  ;

These changes also support utilization of the SFAS shutdown bypass modification (MOD 90-0006) scheduled to be completed during the seventh refueling outage. l C. Safety Assessment and Significant llazards Consideration: Set o Attachment.

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. Docket Number 50-346 License Number NPF-3 Serial Number 1918 httachment Page 1 of 12 SAFETY ASSESSMENT AND SIGNIFICANT llAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NO. 91-0006 L

TITLE I

A proposed change to the Davis-Besse Nuclear Power Station, Unit No. 1 l Operating License, Appendix A, Technical Specification Table .1.3-3 and Table 4.3-2.

DESCRIPTION The puipose of this Safety Assessment and Significant Hazatds Consideration '

is to review the proposed changes to the Davis-Besse Nuclear Power S;ation Unit Number 1 Operating License and Technical Specifications to ensure that the changes do not involve a significant hazard consideration. The following changes to the Technical Specifications are proposed:

a. Revise the Applicable Modes requirements of Table 3.3-3, Item 5.n.

Decay lleat Isolation Valve Interlock channels, to Modes 1,2,3 only (i.e., delete operability requirements while the plant is in Mode 4 or 5),

b. Revise the Applicable Modes requirements for Tahle 3.3-3, Item 5.b, Pressurizer lleaters Interlock channels, to Mode 3 only and further restrict applicability with the footnote, to the period when either one of the two Decay lleat Isolation Valves (011-11 or Dil-12) is open, while in Mode 3,

c. Remove the valvet of Technical Specification 3.0.4 from Table 3.3-3, Item 5.b, Pressurizer lleatet Interlock channels, and,

d. Revise Table 4.3-2, items 5.a and 5.b to have Mode requitements which are consistent with those proposed above.

SYSTEMS, COMPONENTS AND ACTIVITIES AFFECTED The proposed changes affect the Limiting Conditions lot Operation and st .:veillance tes ting t equirements f or the Sa f ety Fea tur es Ac tua tion Sys t em Instrumentation.

SAFETY FUNCTIONS OF Tile AFFECTED SYSTEMS. COMPONENTS AND ACTIVITIES The safety function of the Safety Features Actuation System (SFAS) is to monitor varlous plant non-nuclcat patrmtets and 'o initlate automatic protective actions before plant design limits ate exceeded. One of the SFAS Reactor Coolant System (RCS) ptessure instruments ptavides an input to control circuit of valve Dil-11, which is one of the decay heat temoval I

system isolation valves. This is one of the two valves which separate the I RCS from the DHR system when the RCs pressure is above the design pressuie l

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Docket Number SD-346 License 14 umber NPF-3 Setial !Jumber 1918 Attachment Page 2 of 12 of the DilR system, in the event that the RCS pressure exceeds the Technical Specification set point (<438 psig), while Dil-11 is open, the STAS enables the closing citcuits to Dil-11. If DH-11 is closed when the RCS ptessure exceeds the setpoint, the SFAS disables the opening control circuit of DH-11.

The nafety function of the Reactor Coolent System (RCS) is to provide cooling vater to the reactor fuel dur4 1g power operation and shutdown conditions.

In ordet to pt eserve the integrity of the RCS piessute boundary, oveipressurization protection needs to be provided in all modes of operation.

The safety function of the Decay Heat Removal (D!!R) system is to transfet iission product decay heat and other residual heat itom the teattor so that fuel and RC5 pressure boundaty design limits are not exceeded.

EFFECTS ON SAFETY Duting Modes 4 and 5 opetation the RCS is dit ectly connected to the DilR system via Dil-11 and Dil-12, the Decay llent Isolation Valves. In these Modes, the Lov Temperature Ovetpressurization (LTOP) protection of the RCS is normally provided by the decay heat removal system relief valve (Dil 4849), which is also protecting the DilR system ftem overptessurization.

While the RCS is above ine DilR system design ptessure, diverse iv.erlocks have been provided in the plant design to ptevent DHR ovein._ssurization.

These interlocks, the Decay lleat Isolatirn Valve intetlocks, vill close and prevent opening 011-11 and Dil-12 when the RCS pressut e exceeds the allovable setpoint which is cuttently <438 psig.

While in Modes 4 and 5, closute of DH-ll or Dil-12 vould have two major offeets on the plant. First, the RCS vould be left without LTOP protection.

Second, the DHR pumps vould potentially be operated without a sucti n soutee, which could damage the pumps so that the DHR safety function could be impaired. In tecognition of these effects, TS. 3.4.2 tequites, among other things, that when the plant is in Modes 4 and 5, the control power be temoved from DH-ll and DH-12 after they are opened. Toledo Edison has evaluated the various scenarios which could cause a ptessure transient while in Modes 4 or 5. The r elief valve. Dil-4849. vas sized to accommodate the flov from two liigh Pressure Injec t ion Pumps, which has been determined to be the latnest overpressure source, nnte the RCS is connected to the DHR system via DH-11 and DH-12. DH-4849 vill limit both the PCS and the DilR ptessute to lesr than the DHR design pressure. (Thir inf or ma t ion was previously submitted to the NRC in Toledo Edison Lettet Setial No. 260, dated April 7, 1977.)

Since it is not necennan to b:em nns 1ind of pint <<tive !nterlock to ptotect the DHR System from overpresrotiration fia the RCS while in Mode 4 and Mode 5, the requitement for the SFA5 Decay lleat 1 solation Valve inteilock channel to be operable while in Mode 4 and Mode 5 can be temoved l

. Docket Number 50-346 .

hicense Number NPF-3 Serial Number 1918 Attachment Page 3 of 12 vith no effect on plant safety. Since there is no operability tequirement, there is no need to perform surveillance testing on the instrumentation I

while the plant is in Mode 4 and Mode S.

Branch Technical Position EICSB 3 Isolation of Lov Pressure Systems from the liigh Pressure Reactor Coolant System, has the following applicable guidance:

The following measutes should be incorporated in designs of the interfaces between lov pressure systems and the high ptessute teactor coolant systemt

1. At least two valves in series should be provided to isolate any l

, subsystem whenever the ptimary system pressure is above the pressute 'l rating of the subsystem. I

?. For system interfaces where both valves ate motor operated, the valves should have indepenknt and diverse intetlocks to prevent both from opening unless the primary system pressure is below the ,

subsystem design pressure. Also, the valve opetators should receive '

a signal to close automatically whenever the primary system pressure exceeds the subsystem design pressure. l Dased on this guidance, when the plant is in Modes 1,2, and 3 1t is therefore necessary to have an interlock which prevents DH-ll and DH-12 from being opened until the RCS pressure is belov the DHR system design pressure.

Consequently, the operability and surveillance requirements for the SPAS Decay Heat Isolation Valve 'nterlock are to be maintained for these plant modes.

In order to ensure that double valve protection is established between the RCS and the DHR/Lov Pressure Injection (LPI) system prior to raising the RCS pressure above the DilR system design pressure, interlock channels with the l pressurizer (PZR) heaters have been included in the SFAS design. These interlock channels prevent PZR heater operation if either DH-11 or DH-12 is ofi its closed seat while the RCS pressure is above the interlock's setpoint. These interlock channels vill serve as a automatic prompt to the plant operators to properly position the valves and enable the Decay lleat Isolation Valve interlocks prior to raising RCS pressute. The PZR heater interlock channels need only be operabla in Mode 3 while either of the Decay Heat Isolation Valves is open. Once both valves are closed, the Decay llent *

! Isolation Valve Interlocks provide the redundant. diverse overpressurization l protection of DHR system and the above cited Branch Technical Position guidance is fulfilled.

Vhen cooling down the plant, the Decay Heat Isolation Valve Interlock i prevents opening DH-ll or DH-12 until RCS pressure has been reduced belov l the allovable setpoint. Once one of t.he Decay Heat Isolation Valves is l open, the PZR heatei interlock channel: vill pt event pre: sut e f rom being raised until both DH-11 and DH-12 are open and DH-4049 i t, ptotecting both l

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Docket Number 50-346  !

License Number NPf43 Serial Number 1918 Attachment Page 4 of 12 ,

the HCS and the DilR system. Again, the PZR lleater Interlock need only be opetable in hode 3 vhile either of the Decay Heat Isolation Valves is open.

Therefore, revising the Technical Specification Table 3.3-3 to s equit e operability of the. interlock channels only during this petiod of time has no adverse offeet on safety.

Since the PZR lleater Interlock channels are to be used to protect the DilR

ystem f rom overpressurization it is necessary for it to be operable prior to entering the Mode (or in this case, other specified applicability condition) where it is requited for protection. Consequently the valver of ,

Technical Specification 3.0.4 is being remosed from Table 3.3-4. This change has no adverse effect on plant safety.

Because the Pressurizer lleater Interlock channels are only required to be operable for a limited portion of Mode 3 operations, the Surveillance Requirements for listed in Technical Specification Table 4.3-2, Item 5.b.2 will only be required vhile the plant is in Mode 3 with either of the Decay i

Heat Isolation Valves open, consistent with the Limiting condition for ope.ation requitement of Table 3.3-3. This requirement and Technical Specification 4.0.4 vill ensure that the interlock channels are verified to be operable prior to and duting intervals when they are being utilized to i protect the DHR system from overpressurization. Therefote, this change has no adverse effect on safety.

SIGNIFICANT llAZARDS CONSIDERATION

,' The Nuclear Regulatory Commission has provided standards in 10CFR50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License (or a facility. A proposed amendment i involves no significant hazards consideration if operation of the facility

.in accordance with the proposed changes vouldt (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 accident previously evaluated; or (3) Not involve a significant reduction in a margin of salety. Toledo Edison had reviewed the proposed change and determined that a significant hazards consideration.

does not exist because operation of the Davis-Besse Nuclear Power Station, Unit 1 in accot ence vith these changes vould:

la. Not involve a significant increase in the probability of an accident previously evaluated because the same level of protection la being g provided as before.

Ib. Not involve a significant increase in the consequences of an r accident previously evaluated because the plant response to an I. overptessure condition vill be unchanged.

2a. Not create the possibility of a new kind of accident from an accident previously evaluated because no new failure modes are inttoduced, and, therefote, no new accident scena ios can be postulated.

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, Docket Number 50-346 i License Number NPF-3 Serial Number 1918 l Attachment Page 5 of 12

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l 2b. Not create the possibility of a different kind of accident from any accident previously evaluated because there are no dif'.erent accidents postulated from those previously evaluated.

3. Not involve a significant reduction in a matgin of safety since the ,

overpressute protection of the DHR system and the '(CS vill be preserved.

CONCLlfS10N Based on the above Toledo Edison has determined that this License Amendment Request does not involve a significant hazards consideiation. As this License Amendment Request concerns a proposed change to the Technical ,

Specifications that must be revieved by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety -

question.

ATTACllMENT I

Attached is the proposed mark-up change to the Operating License.

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