ML20207H919

Amend 117 to License NPF-3,revising Tech Spec Table 4.7-1 & Section 3/4.7.1.1 Re Main Steam Safety Valve Setpoints & ASME Code Requirements

ML20207H919
Person / Time
Site:	Davis Besse
Issue date:	08/24/1988
From:	Perkins K Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207H923	List: ML20207H919 ML20207H927 ML20207H931
References
NUDOCS 8808300114
Download: ML20207H919 (11)

Text

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8 NUCLEAR REGULATORY COMMISSION o

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TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendinent No.117 License No. NPF-3 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by the Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated March 4, 1988 as supplemented May 4,1988 complies with the standards and requirementsoftheAtomicEnergyActof1954,asamended(theAct),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assuranco (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will r.ot be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. NPF-3 is hereby amended to read as follows:

I 8008300114 000824 DR ADOCK 050 6

l.

2 (a) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.117, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented not later than October 7,1988.

FOR THE NUCLEAR REGULATORY C0fmISS10N f

k Kenneth E. Perkins, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V, & Special Projects Attachn. ant: Changes to the Technical i

Specifications Date of Issuance: August 24, 1988 l

'3/4.7 PLANT SYSTEMS y

3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE.

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APPLICABILITY: MODES 1, 2 ana 3.

'N.

ACTION:

With one or more main steam line code safety valves inoperable, operation in HODES 1, 2 and 3 may proceed provided, that within a.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either,

'the inoperable valve is restored to OPERABLE status, or, 1.

2.

a) the High Flux Trip Setpoint is reduced per Table 3.7-1,

and, b) there are a minimum of two OPERABLE safety valves per steam generator, at least one with a setpoint act greater than 1050 psig (+/-1%)*, and, c) no OPERABLE safety valve has a setpoint greater than 1100 psig (+/-1%)*;

otherwise, b.

be in at least 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 COLD 1

SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, The provisions of Specification 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5, are applicable for the main steam line code safety valves of Table 4.7-1.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

DAVIS-BESSE, UNIT 1 3/4 7-1 Amendment No. 117

TABLE 3.7-1 i

E MAXIMUM ALLOWABLE HIGH FLUX TRIP SETPOINT WITH INOPERAELE 1

Y STEAM LINE SAFETY VALVES M

Maxianum Allowable g

Mariana Number of Inoperable Safety High Flux Trip Setpoint*

l Valves on Any Steam Generator (Pe'rcent of RATED THERMAL POWER)

I 0.95W 2

0.82W 3

0.69W 4

0.56W R

5 0.43W e~

y 6

0.31W w

7 0.18W tb

.f l

• Based on High Flux Trip Setpoint for four pump operation, W, as per Table 2.2-1.

D.

W e

G O

9 e

r--er-.

- w w-w

. = ~ - -. -

ts p

-m 9

E w

Table 4.7-1 N

E MAIN STEAM LINE SAFETY VALVE LIFT SETTINGS

+4 NUMBER PER STEAM GENERATOR LIFT SETTING (i 1%)*

a.

2 1050 psig 1

l b.

7 1100 psig u

4 k

t 4

w I

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• The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

n t

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"J ta j

b 8

i W

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PLANT SYSTEMS AUXILIARY TEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Two independent steam generator auxiliary feedwater pumps and associated flow paths hhall be OPEkABLE.

APPLICABILITY: NODES 1, 2 and 3*.

ACTION:

With one Auxiliary Feedwater System inoperable, restore the a.

inoperable system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each Auxiliary Feedwater System shall be demonstrated OPERABII:

At least once per 31 days on a STAGGERED TEST BASIS by:

a.

l 1.

Verifying that each steam turbine driven pump develops a differential pressure of > 1070 psid on recircul.ation flow when the secondary steam supply pressure is greater than 800 psia, as ocasured on PI SP 12B for pump 1-1 and PI SP 12A for pump 1-2.

2.

Verifyi.g that each v'alve (power operated or actonatic) in the flow path is in its correct position.

3.

Verifying that all manual valves in the auxiliary feadwater pump suction and discharge lines that affect the system's capacity to deliver water to the steam generator are locked in their proper position.

b.

At least once per 18 months by:

1.

Verifying that each automatic valve in the flow path 1

actuates to its correct position on an auxiliary feedwater actuation test signal prior to entering HODE 3.

• 2.

Verifying that each pump starts automatically upon receipt of an auxiliary feedwater actuation test signal prior to entering HODE 2.

3.

Verifying that there is a flow path between each auxiliary feedwater pump and each steam generator by pumping water from the Condensate Storage Tank to the steam generator.

• Provision of section 3.0.4 is not applicable for. entry into MODE 3.

DAVIS-BESSE, UNIT 1 3/4 7-4 Amendment No. 96

ql 3/4.7 ptANT SYSTEMS BASES 3/4.7.1 TVRBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary systes pressure will be limited to within 110% its design pressure of 1050 psis during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The safety valve set pressures and relieving capacities are in accordance with Section III of ASME Boiler and Pressure Vessel Code, 1971 Edition.

The Code requires the following:

1.

At le*ast two pressure-relief valves are required to provide relieving capacity for steam systems.

2.

The capacity of the smallest pressure-relief valve shall not be less thas %0 percent of that of the largest pressure-relief device.

3.

The set pressure of one of the pressure-relief devices shall not be greater than the maximus allowable working pressure of the systes at design temperature.

4.

Total rated relieving capacity of the pressure-relief devices shall preven.: a rise in pressure of more than 10 percent above systes design pressure at design temperature under any pressure transients anticipated to arise.

These requirements are, respectively, set as follows:

1.

Nine safety valves are installed per steam generator.

2.

The relief capacity of two of the nine safety valves per steam generator is 583,574 lbs/br each, and the capacity of the remaining seven is 845,759 lbs/hr each.

3.

A minimum of two OPERABLE safety valves per steam generator, with a combined total relief capacity of at least 1,167,148 lbs/br, one with a setpoint not greater than 1050 psig (+/-l%),

and one with a setpoint not greater than 1100 psig (+/-1%).

4.

The total relieving capacity of all safety valves on both main steam lines is 14,175,000 lbs/hr which is 120 percent of the total secondary system flow of 11.760,000 lbs/br at 100 percent of rated thermal power. A maximum safety valve setpoint pressure of 1100 psig (+/-1%) assures main steam systes pressure remains below 110 percent, or 1155 psig.

DAVIS-BESSE, UNIT 1 B 3/4 7-1 Amendment No. 48.117

3/4.7 PLANT SYSTEMS BASES 3/4.7.1.1 SAFETY VALVES (Cantinued) r STARTUP and/or F0WER OPERATION is allowable with safety val,e inoperable within the limit ations of the ACTION requirements on the basis of the l

reduction in secondary system steas flow and THERMAL POWER required by the reduced reactor trip settings of the High Flux channels. The reactor trip

(

setpoint reductions are derived on the following bases:

3p, (X) - (Y)(V), y where:

SP =, reduced Trip Setpoint in percent of RATED TFIRMAL POWER (Not to Exceed W)

V = maximum number of inoperable safety valves per steam generator W = High Flux Trip Setpoint for four pump operation as spect-fied in Table 2.2-1 X = Total relieving capacity of all safety valves per steam generator in Ibs/ hour, 7,087,500 lbs/ hour Y = Maximum rel$rvit:g capacity of any one safety valve in lbs/bour, 045,.759 3ds/nour 5

2 = Required reliev;ng ecpacity per steam generator in lbs/br, 6,585,600 lbs/br.

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Next page is B 3/4 7-2 DAVIS-BESSE, UNIT 1 B 3/4 7-la Amendment No. 117 l

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PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY TEEDWATER SYSTEMS The OPERABILITY of the Auxiliary Feedwater Systems ensures that the Reactor Coolant System can be cooled down to less than 280'F from normal operating conditions in the event of a total loss of offsite power.

Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 600 sps at a pressure of 1050 psis to the entrance of l

the steam generators. This capacity is sufficient to ensure that adequate feedvater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 280'T where the Decay Heat Removal System may be placed in operation.

Following any modifications or repairs to the Auxiliary Feedwater System piping from the Condensate Storage Tank through auxiliary feed pumps to the steam generators that could affect the system's capability to deliver water to the steam generators, following extended cold shutdown, a flow path verification test shall be performed.

This test may be conducted in MODES 4, 5 or 6 using auxiliary steam to drive the auxiliary feed pumps turbine to demonstrate that the flow path exists from the Condensate Storage Tank to the steam generators via auxiliary feed pumps.

3/4.7.1.3 CONDENSATE STORAGE FACILITIES The OPERABILITY of the Condensate Storage Tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> with steam discharge to atmosphere and to cooldown the Reactor Coolant System to less than 280'T in the event of a total loss of offsite power or of the main feedwater system. The contained water volume limit ideludes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secoudary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a j

steam line rupture. This restriction is required to 1) aintaise the i

1 DAVIS-BESSE, UNIT 1 B 3/4 7-2 Amendment No. 58, 96, 117 j

Rev. 11/18/87 1

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