ML20138M125

Application for Amend to License NPF-3,requesting Delay of Heat Removal Sys Valve Pit Surveillance Requirements

ML20138M125
Person / Time
Site:	Davis Besse
Issue date:	02/14/1997
From:	Jeffery Wood CENTERIOR ENERGY
To:
Shared Package
ML20138M120	List: ML20138M116 ML20138M125 ML20138M128
References
NUDOCS 9702250271
Download: ML20138M125 (7)

Text

. _ - . _ . . _ ..m . - - . . - - - _ _ . . _ .

- . . _ _ _ _ . _ _ . _ - . _ . . . _ . _ ._ . _ _ ~ . - - . . = _ .

Dockst Number 50-346 .

Licanoa Number NPF-3  !

f ", . Sarial Numb 3r 2448  !

i Enclosure l ,Page 1  ;

i l

APPLICATION FOR AMENDMENT l

TO l

\ t FACILITY OPERATING LICENSE NUMBER NPF-3 i t

l DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 i

Attached is the requested change to the Davis-Besse Nuclear Power Station, f Unit Number 1 Facility Operating License Number NPF-3. Also included is ,

the Safety Assessment and Significant Hazards Consideration.  ;

The proposed change (submitted under cover letter Serial Number 2448) )

concern: -

Appendix A, Technical Specification Section 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - T >280*F, and associated Bases 3/4.5.2 and 3/4.5.3, ECCS Subsystems. # 9 For: J. K. Wood, Vice President - Nuclear By:  ! AJ -

T. J. rs, Director - Nuclear Assurance i

Sworn to and subscribed before me this 14th day of February, 1997. l l

h Notary Public, State of Ohio  ;

LAURA A.JDe#50N Notary Public, State of Ohio My Commis' ion Empires 8 15-2001 f-l 9702250271 970214 PDR ADOCK 05000346:

P PDR

, Dockst Number 50-346

. Licanse Number NPF-3 Serial Number 2448

. Attachment 1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 97-0006 (12 pages follow) l I

{

f r

1 l

)

l 1

I . .

LAR 97-0006 Page i SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 97-0006 i

l l

TITLE 1 A Proposed Change to Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A, Technical Specification (TS) 3/4.5.2, Emergency Core Cool-ing Systems - ECCS Subsystems - T 2280*F, and Associated Bases 3/4.5.2 and 3/4.5.3, ECCS Subsystems.

1 DESCRIPTION:

l The purpose of this Safety Assessment and Significant Hazards Consideration is to review the proposed change to the Davis-Besse Nuclear Power Station (DBNPS),

Unit No. 1 Technical Specifications to ensure that the change does not involve a significant hazard consideration.

Technical Specification (TS) 3/4.5.2 Surveillance Requirement (SR) 4.5.2.f currently requires that a vacuum leakage rate test of the watertight enclosure for the motor operators of valves DH-ll and DH-12 in the Decay Heat Removal System drop line be performed. The test ensures that the motor operators on 1 valves DH-11 and DH-12 will not be flooded for at least 7 days following a l Loss-of-Coolant Accident (LOCA). These motor operators, located in a pit, are  !'

not qualified for the submerged environment which would be experienced post-LOCA. Flooding could cause the motor operators to fail and the valves to remain in their closed position. The valves are required to be opened post-LOCA. The  !

watertight enclosure consists of the walls of the valve pit and large 1/4-inch I deck plates attached to a steel frame which covers the valve pit. The test is required to be performed: 1) At least once per 18 months; 2) After each opening of the watertight enclosure; and 3) After any maintenance on or modification to the watertight enclosure which could affect its integrity.

The proposed change would clarify TS SR 4.5.2.f by adding a paragraph to read:

The inspection port on the watertight enclosure may be opened without requiring performance of the vacuum leakage rate test, to perform inspections. After use, the inspection port must be verified as closed in its correct position. Provisions of TS 3.0.3 are not applicable during these inspections.

Associated with this change, Bases 3/4.5.2 and 3/4.5.3 would be revised by adding a paragraph to read as follows:

Decay Heat Removal System valves DH-ll and DH-12 are located in an area that would be flooded following a LOCA. These valves are located in a watertight enclosure to ensure their operability up to seven days following a LOCA. Surveillance Requirements are provided to verify the acceptable leak tightness of this enclosure. An inspection port is located on this watertight enclosure, which is typically used for performing inspections inside the enclosure. During the vacuum leakage rate test, the inspection 1

.. u ~> >>_.z-~- ~ ~ . . - . - a .w, = usx - + . . -

[. -

. i LAR 97-0006 .

j i

Page 2 i

port is in a closed position and subject to the test. The inspection port may be subsequently opened for use in viewing inside the enclosure.  !

Opening this inspection port will not require performance of the vacuum leakage rate test because of the design of the closure fitting, which will preclude leakage under LOCA conditions, when properly installed. Proper l installation includes independent verification.

I l SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED: -

Low Pressure Injection System.

Valves DH-11 and DH-12, including their watertight enclosure and inspection port. l Vacuum leakage rate testing of the watertight enclosure.

J l

l FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS AND ACTIVITIES:

1 The Low Pressure Injection (LPI) system provides a source of borated water directly to the reactor vessel following a large break Loss-of-Coolant Accident (LOCA). It also is used to provide long term core cooling by recirculating water from the containment emergency sump to the core. It can be used to supply borated water to the suction of the High Pressure Injection Pumps, to supply auxiliary spray to the pressurizer and to provide long term boron dilution for the reactor vessel. In its normal mode of operation, the system is used to remove heat from the Reactor Coolant System (RCS) and decay heat from the core during plant cooldowns and shutdowns.

The safety function of Decay Heat Removal (DHR) isolation valves DH-11 and DH-12 is to isolate the RCS from the DHR System when the RCS pressure is greater than the pressure rating of the DHR System. The safety function of DHR isolation valves DH-11 and DH-12, relating to this proposed License Amendment Request (LAR), is also to provide a circulation flow path to prevent boron concentration build-up and boron precipitation in the core post-LOCA, as described in Updated Safety Analysis Report (USAR) Section 6.3.3.1.2, "Results of Analysis (Large Break)."

The function of the watertight enclosure for the DH-11 and DH-12 valve pit is to ensure that the motor operators on these valves will not be flooded within seven days following a LOCA. l l

The function of the vacuum leakage rate testing activities is to ensure the l watertight enclosure is capable of performing its function. Surveillance  !

Requirement 4.5.2.f is a non-standard, plant-specific requirement that was added I to the.DbNPS Technical Specifications at the time the Operating License was issued in 1977. This requirement was added due to the design of the valve pit cover. The inspection port was later added in 1986.

l The decay heat valve pit inspection port consists of a 4-inch (nominal) pipe stub welded to the pit cover and sealed with a 4-inch (nominal) adapter and cap

("Kamlok" coupling) arrangement manufactured by Dover Corporation /OPW Division.

The Kamlok coupling is designed to be removed and reinstalled, and contains an  ;

integral viton gasket for sealing. The Kamlok coupling, which is designed for l 380 deg-F temperature and 150 psig pressure, inherently preclude- leakage under the possible range of conditions, when properly installed.

l LAR 97-0006 li[ Page 3 The Kamlok coupling also allows personnel to quickly and easily access the

inspection port, and to quickly and easily close the inspection port upon completion of inspection activities.

EFFECTS ON SAFETY:

The sealing capability of the Kamlok coupling is confirmed during performance of the SR 4.5.2.f watertight enclosure vacuum leakage rate test. No problems relative to the sealing performance of the Kamlok coupling have been observed during tests performed since its installation.

The surveillance test for SR 4.5.2.f is designed to be performed in Modes 4 (Hot Shutdown), 5 (Cold Shutdown), or 6 (Refueling). The surveillance test is typically performed in Mode 5 (Cold Shutdown) since it is desirable to avoid extended plant operation in Mode 4 due to Reactor Coolant Pump net positive suction head considerations. The surveillance test is not performed during Modes 1 (Power Operation), 2 (Startup) or 3 (Hot Standby) due to the potentially large differential pressure which would exist across the watertight pit cover with the pit vent isolated, the pit under vacuum conditions during the conduct of the test, and the high containment vessel pressure which would exist were a LOCA to occur during the test. In addition, there are personnel dose and heat .

stress concerns which would make an extended stay in containment for performance of the test difficult.

This license amendment request would allow the inspection port to be opened to perform inspections inside the enclosure, without requiring performance of the vacuum leakage rate test upon closure. Immediately after use, the proposed change would require the inspection port to be verified closed in its correct position.

Based on the above, there is reasonable assurance that the decay heat valve pit inspection port, as designed and installed, is leak-tight and will not adversely affect the decay heat valve pit watertight enclosure sealing capability.

Consequently, there will be no adverse effect on plant safety.

This license amendment request would also revise SR 4.5.2.f by stating that the provisions of TS 3.0.3 are not applicable during these inspections. Entry into TS 3.0.3 while the inspection port is opened to conduct an inspection, in a plant condition in which LCO 3.5.2 applies (Modes 1 through 3), would impose the i significant and unwarranted action on the DBNPS of reporting the event under the requirements of 10 CFR 50.72 and 50.73, and preparing for an expedited plant shutdown.

The inspection port is presently opened in Mode 3 following each refueling outage in order to insert a camera for performance of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code visual inspection  ;

requirements for portions of the decay heat piping located within the pit. This '

inspection requires at least Mode 3 plant conditions for pressure and temperature requirements. There is also the potential need to open the inspection port to confirm, by inspection, level switch indication of leakage within the pit. The inspection port is also opened during plant heatup in accordance with procedure DB-OP-06900, " Plant Heatup," to visually check for leakage.

l 1

._ _ . . _ _ __ . . _ . . _ _ _ . _ . _ _ . ___ _ _ __ __ .__ m..

LAR 97,0006 Page 4 Based on the above, the proposed change regarding the provisions of TS 3.0.3 will have no adverse effect on plant safety.

The proposed Bases change is associated with the proposed cnanges to SR 4.5.2.f.  !

This Bases change explains the reason for the Surveillance Requirement on the watertight enclosure and the related requirements relative to the inspection port. Therefore, this change will have no adverse effect on plant safety. l' SIGNIFICANT HAZARDS CONSIDERATION: i 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 involves no signifi-  !

cant hazards consideration if operation of the facility in accordance with the ,

l proposed changes would: (1) Not involve a significant increase _in the proba- i bility 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 safety.

Toledo Edison had reviewed the proposed change and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nu-clear Power Station (DBNPS), Unit 1 in accordance with these changes would:

i la. Not involve a significant increase in the probability of an accident pre-I viously evaluated because the initiators regarding the large break Loss-of-Coolant Accident (LOCA) are not affected by the proposed change. Revising ,

Surveillance Requirement 4.5.2.f has no bearing on initiating an accident previously evaluated. The flow path through the decay heat drop line also is not an accident initiator. l lb. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed change does not alter the source term, containment isolation, allowable radiological releases, or invalidate the assumptions used in evaluating radiological releases. Therefore, the radiological consequences of all accidents presented in the DBNPS Updated Safety Analysis Report (USAR) are unchanged.

2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because the operability requirements of Decay Heat Removal (DHR) System isolation valves DH-ll and DH-12 will continue to be adequately addressed by Surveillance Requirement 4.5.2.f. The plant will be operated in the same way as before, and no different accident initiators or failure mechanism are introduced by the proposed change. The inspection i port's Kamlok coupling is included as part of the watertight enclosure vacuum leakage rate test to ensure its leak tightness. In addition, the proposed change adds a new stipulation to Surveillance Requirement 4.5.2.f i that after its use, the inspection port must be verified as closed in its correct position. Thus, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Not involve a significant reduction in a margin of safety because the l

proposed change does not involve any new changes to the initial conditions contributing to accident severity or consequences. The inspection port's Kamlok coupling is included as part of the watertight enclosure vacuum

l

. I i

l LAR 97-0006 l Page 5 leakage rate test to ensure its leak tightness. In addition, the proposed I change adds a new stipulation to Surveillance Requirement 4.5.2.f that after its use, the inspection port must be verified as closed in its correct j position. The design of the Kamlok coupling provides for quick and easy l access to the inspection port, and quick and easy closure of the inspection  !

port upon completion of inspection activities. Consequently there are no l reductions in a margin of safety. '

CONCLUSION:

On the basis of the above, Toledo Edison has determined that the License Amend-ment Request does not involve a significant hazards consideration. l ATTACHMENT:

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

REFERENCES:

1. Davis-Besse Nuclear Power Station Operating License NPF-3 Appendix A Technical Specifications through Amendment 212.

2. Davis-Besse Nuclear Power Station Updated Safety Analysis Report (USAR) )

through Revision 20. )

l l

l j