ML20205H409

Application for Amend to License NPF-3,reducing Auxiliary Feedwater (AFW) Flow Requirement for Steam Driven AFW Pumps from 800 to 600 Gpm,Per B&W Analysis of Flow & Per 861117 Commitment.Fee Paid

ML20205H409
Person / Time
Site:	Davis Besse
Issue date:	03/23/1987
From:	Shelton D TOLEDO EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205H413	List: ML20205H409 ML20205H453
References
1360, TAC-65068, NUDOCS 8704010046
Download: ML20205H409 (9)

Text

r TOLEDO Docket No. 50-346 EDISON License No. NPF-3 Serial No. 1360 DONALD C. SHELTON March 23, 1987 pg%%*

United States Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555 Gentlemen:

The purpose of this letter is to request a change to the Bases of the Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Technical Specifications. The proposed changes, included as an attachment to this letter, involve Bases Section 3/4.7.1.2, Auxiliary Feedwater Systems.

The proposed changes are being submitted in accordance with Toledo Edison's commitment in its letter to the NRC, dated November 17, 1986 (Serial No.

1322), to propose a request to reduce the Auxiliary Feedwater (AFW) flow requirement in the AFW Technical Specification Bases Section 3/4.7.1.2, within 90 days after restart of Davis-Besse.

As discussed in the attached Safety Evaluation, the proposed change involves reducing the AFW flow requirement for the steam driven AFW pumps from 800 gpm to 600 gpm based on a recent analysis of AFW flow performed by the Babcock and Wilcox Company for Toledo Edison. This plant-specific analysis demonstrated that adequate cooling was achieved, for a loss of feedwater event, with a nominal AFW flow of 600 gpm at 1050 psig.

The Bases change requested herein does not require the submittal of a License Amendment Request based on the provisions of 10CFR50.36(a), which requires inclusion of Bases for Technical Specifications but indicates such Bases shall not become part of the Technical Specifications.

Toledo Edison requests that the NRC approve this change by May 18, 1987, to be effective upon NRC approval.

Enclosed is a check for $150 as required by 10CFR170.12(c) for an appli-cation fee.

Very truly y,urs, I ,

DCS:JAE: CAB:p1f dr k

Attachments tii 0 0\

GhUgD1 cc: DB-1 NRC Resident Inspector M State of Ohio g\

A. B. Davis, Acting Regional Administrator (2 copies)

A. W. DeAgazio, NRC/NRR Davis-Besse Project Manager THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652 87040;10046 870323 PDR ADOCK 05000346 P PDR

D:ck:t No. 50-346 License No. NPF-3 Serial No. 1360 Attachment 1 Page 1 SAFETY EVALUATION DESCRIPTION OF PROPOSED ACTIVITIES This Bases change request, as committed to by letter dated November 17, 1986 (Serial No. 1322) from J. Williams, Jr. (Toledo Edison) to J. F. Stolz (NRC), is to change Technical Specification Bases 3/4.7.1.2 pertaining to the Auxiliary Feedwater (AFW) System. This request proposes lowering the AFW flow requirement for each steam driven Auxiliary Feedwater Pump (AFP) from 800 gpm to 600 gpm at a pressure of 1050 psig at the entrance to the steam generators. This request is a result of recent ana!ysis performed by the Babcock and Wilcox Company (B&W) for Toledo Edison.

SYSTEMS AFFECTED The AFW System consists of two steam turbine-driven feedwater pumps, condensate storage tanks, suction and discharge water piping, steam piping, valves and associated instrumentation and controls. The pumps can take suction from the condensate storage tanks, the fire protection system, and from the Class I service water system. The condensate storage capacity is sized so that the minimum total condensate inventory available to the pumps is sufficient to remove decay heat for approx-imately thirteen hours plus a subsequent cooldown to 280*F. Following a complete loes of normal and reserve power the AFW System supplies water directly to the steam generators through the AFW nozzles to remove reactor decay heat. Reactor decay heat removal in the absence of the reactor coolant pumps is provided by the natural circulation characteristics of cl.e Reactor Coolant System (RCS). Use of the AFW System for cooldown is discontinued when the RCS temperature decreases to about 280*F; further cooldown is accomplished by the Decay Heat Removal System.

Reducing the AFW flow requirement does not alter the present configuration of the s" stem. Analysis, as described below, ensures that this change maintains an adequate AFW flow to remove core decay heat. Therefore, lowering the AFW flow does not increase the demand placed on any system.

DOCUMENTS AFFECTED

1. Technical Specification Bases 3/4.7.1.2

2. USAR Section 9.2.7

3. USAR Section 15.2.8 SAFETY FUNCTION OF SYSTEM AFFECTED The safety function of the AFW System is to provide feedwater to the steam generators for the removal of reactor decay heat in the absence of main feedwater and to promote natural circulation of the RCS in the event of a loss of all four reactor coolant pumps.

I Docket No. 50-346 License No. NPF-3 Serial No. 1360 Attachment 1 Page 2 EFFECTS ON SAFETY The existing Bases states that "each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 800 gpm at a pressure of 1050 psig to the entrance of the steam generators". The Basis goes on to say that "this capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 280*F...". This latter quote also defines the function of the AFW System. This definition of function is unaffected by the lower flow requirement.

The performance of the AFW System is described in the USAR Chapter 15 accident analyses for the Loss of Feedwater Transient (Section 15.2.8).

This transient puts more severe design requirements on the AFW System than other transients. The USAR analysis used an AFW flowrate of 800 gpm to be delivered within 40 seconds of actuation of the AFW System.

The analysis performed by B&W in support of reducing the flow requirement, B&W document 32-1159090-01, is the Loss of Feedwater Transient analysis described in USAR Section 15.2.8. The B&W analysis includes the following assumptions:

1. Initial reactor power is 102% Full Power

2. No credit for PORV, pressurizer sprays, or make-up flows

- 3. 1.2 times ANS 5.1 (1979) decay heat curve

4. Full AFW flow delivered to the steam generator 40 seconds after Steam and Feedwater Rupture Control System (SFRCS) low level setpoint is reached

5. SFRCS low level setpoint of 10 inches actual level above lower tubesheet

6. Offsite power is available during event. Therefore, RCS pumps continue to operate

7. Turbine trip due to reactor trip Consistent with the USAR Chapter 15 acceptance criteria, the acceptance criteria for this B&W analysis are:

1. No fuel damage

2. RCS pressure does not exceed 2750 psig In this analysis, the initiating event is a failure of the main feedwater control valves with a subsequent ramp reduction to zero flow in seven seconds. During the feedwater reduction, the temperature and pressure in the RCS begins to increase and continues until the high pressure trip setpoint of 2400 psia is reached and the reactor trips at approximately 14 seconds into the event. RCS pressure is then controlled by the pressurizer safety valves. This ar.alysis initiates the turbine trip one second following the reactor trip with the turbine stop valves ramping closed during the next second for a total of two seconds from reactor trip to valve closure. At this point, the secondary side pressure is maintained by the main steam safety valves while the steam generator

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Dockat No. 50-346 License No. NPF-3 Serial No. 1360 Attachment 1 Page 3 inventory is boiling down. When the low level setpoint of 10 inches (collapsed liquid level) in the steam generator is reached, SFRCS is initiated. This occurs at approximately 25.6 seconds into the event.

Full flow AFW reaches the steam generator 40 seconds later with AFW feeding only one steam generator due to single failure considerations.

The AFW flow curve used, based on previous analyses, provides a flow of 600 gpm at a 1050 psig steam generator pressure.

As can be seen from Figures A and B respectively, by 170 seconds into the event a sufficient water level is established in the steam generator to terminate the temperature rise in the RCS and to establish RCS cooldown.

In other words, the AFW flow is sufficient for the removal of core decay heat and subsequent cooldown to 280*F. Since the RCS remains subcooled throughout the event with no temperature or power excursions which approach the 112% full power limit, there is no fuel damage. Figure C shows the peak RCS pressure was limited to approximately 2590 psia which is below the safety limit of 110% of design pressure (2750 psig). There-fore, the USAR Chapter 15 acceptance criteria have been met. Figure D shows the maximum pressurizer level reached was approximately 350 inches as measured from a zero point 30 inches above the top of the tap. The pressurizer level did exceed the high end of the scale. However, the pressurizer maintained a steam bubble at all times during the event, (approximately four vertical feet remaining). This analysis is conserva-tive in that the modeling techniques do not consider any cooling of the RCS due to AFW impinging on the steam generator tubes.

Since the Loss of Feedwater is a bounding transient for the AFW System requirements and has been analyzed with acceptable results, the conse-quences of all other transients requiring AFW are also acceptable.

UNREVIEWED SAFETY QUESTION EVALUATION The proposed action would not increase the probability of an accident previously evaluated in the USAR because there have been no design changes (10CFR50.59(a)(2)(1)) .

The proposed action would not increase the consequences of an accident previously evaluated in the USAR because the analysis, performed for the Loss of Feedwater Transient, has shown the reduced flow capable of removing decay heat and meeting the USAR Chapter 15 acceptance criteria.

Thus, although the results are different, the consequences are acceptable (I OCFR50.59(a) (2) (1)) .

The proposed action would not increase the probability of a malfunction of equipment important to safety because there have been no design changes (10CFR50.59(a)(2)(1)).

The proposed action would not increase the consequences of a malfunction of equipment important to safety because analysis has shown, that with different but acceptable consequences, the lower flow is capable of removing decay heat and cooling the reactor core (10CFR50.59(a)(2)(i)).

- .~ ~ . .- - .. .. .- . .- .

Docket No. 50-34'6 4

. License No.

NPF-3 Serial No. 1360 Attachment 1 Page 4

} The proposed cetion would not create a possibility for an accident of a ,

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different type than any evaluated previously in the USAR because there:

have been no design changes and analysis has shown the lower flow capable of removing decay heat and cooling the core (10CFR50.59(a)(2)(ii)).

i The proposed action would not create a possibility for a malfunction of equipment of a different type than any evaluated previously in the USAR i because there have been no design changes (10CFR50.59(a)(2)(ii)).

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The proposed action would not reduce the margin of safety as defined in

< the basis for the Technical Specification. Lowering the AFW flow does 1 increase the RCS temperature and pressurizer level over.that previously analyzed for this specific event but the results meet the USAR Chapter 15 l_ acceptance criteria and are bounded by other Chapter.15 analysis (rod i withdrawal accident for pressure and LOCA for fuel temperature). There-

! . fore, lowering the AFW flow does not reduce the overall margin of safety I as analyzed for thCs plant and consequently does not impact the intent of the Bases (10CFR50.59(a)(2)(iii)).

CONCLUSION Pursuant to the above, this change to Technical Specification Bases

3/4.7.1.2 does not change the intent of the Bases and does not involve an.

! unreviewed safety question by Toledo Edison. In addition, the NRC has l previously reviewed the issue of a reduced AFW flow requirement for other

{ B&W reactors and granted AFW flow reductions.- Therefore, this change does not involve an unreviewed safety question.

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