ML20096C682

Proposed Tech Specs Reflecting Changes in Reactor Vessel Matls Irradiation Surveillance Schedule & Flow Verification of Auxiliary Feedwater Sys

ML20096C682
Person / Time
Site:	Davis Besse
Issue date:	08/27/1984
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20096C643	List: ML20096C640 ML20096C663 ML20096C682
References
TAC-55652, TAC-55782, NUDOCS 8409050375
Download: ML20096C682 (8)

Text

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REACTOR COOLANT SYSTEM .

3/4.4.9 PRESSURE /TEW ERATURE LIMITS REACTOR COOLANT SYSTEM L

LIMITING CONDITION FOR OPERATION

3. 4. 9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3 and 3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100'F in any one hour period, and

b. A maximum cooldown of 100*F in any one hour period, .

f APPLICABILITY: At all times. .

ACTION:

I With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perfonn an engineering evaluation to determine the effects of the out-of-limit condition on the ~

fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation

, or 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 reduce RCS T and pressure to less than 200*F and 500 psig, respectively, within tM9following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS l

1 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inse."vice leak and hydrostatic testing operations.

4.4.9.1.2 The_ reactor vessel material irradiation survei_llance specimens j representative of the vessel materials shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.

8409050375 840827 DR ADOCK 0500034 ,

DAVIS-BESSE, UNIT 1 3/4 4-24

g TABLE 4.4-5 *

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U REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE Nl . .

m Sequence Time of Withdrawal l .

E Il First Earliest of: 1.5 EFPY; capsule fluence > 5 x 10 18 n/cm2 ;

H highest RT NDT f an encapsulated material equals 50F.

3 . +

Second Earliest of

3 EFPY; capsule fluence midway between that

! of the first and third capsules.

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! Third Earliest of: 6 EFPY; capsule fluence corresponds to that of the EOL fluence of the reactor vessel 1/4T location. ..

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• - Fourth Earliest of: 15 EFPY; capsule fluence corresponds to that of the EOL fluence of the reactor vessel-inside sur-g face location.

! Fifth not less than once nor greater than.twice the

} Standby;'nce EOL Flue of the reactor vessel inside surface loca- '

j tion. Capsule may be held without testing after with-j drawal. '

i l Sixth Standby capsule.

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! IRE CTOR C00UWT SYSTEM .

8ASES l The number of reactor vessel 4rradiation surveillance specimens and i the frequencies for removing and testing these specimens are provided in Table 4.4-5.t.The withdrawal schedule is based on four considerations:

(a) uncover possible technical anomalies as early in life as they can be detected (and of first fuel cycle), (b) define the material properties needed to perform the analysis required by Appendix G to 10 CFR 50, (c) reserve two capsules for evaluation of the effectiveness of thermal annealing in the event the inplace annealing becomes necessary (d) provide material property data corresponding to the reactor vessel *

.line surface conditions at the end of Service. The withdrawal schedule of Table 4.4-5 is specified to assure compliance with the requirements of Appendix H to 10 CFR 50. Appendix H references the requirements of .

ASTM E185 for surveillance program criteria. Table 4.4-5 is designed to meet the requirements of ASTM E185-82.

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l Docket No.-50-346 License No. NPF-3 Serial'Ko, 1074 August 27, 1984 Attachment II

1. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 4.7.1.2.

A. Time required to Implement . This change is to be effective upon VRC approval.

B. Reason for Change (Facility Change Request 83-032). In response to NRC request dated February 21, 1984 on the AFW System.

C. Safety Evaluation (See Attached)

D. Significant Hazard Consideration (See Attached)

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Safety Evaluation This amendment request to the Technical Specifications Surveillance requirement Section 4.7.1.2 adds flow. test requirement to. verify Auxiliary Feedwater (AFW) System Operability following: 1) refueling outage (every

-18 months); 2) an extended outage (>30 days in Mode 5); and 3) modification to the system that could affect system capability. It also requires verification of locked valve positions.

The safety function of the Technical Specification surveillance requirement 4.7.1.2 is to insure that the AFW System availability is maintained during Modes 1,'2 and 3 to provide a heat sink to the Reactor Coolant System (RCS) when main feedwater is lost.

The proposed change to-Section 4.7.1.2 will provide adequate assurance that the AFW flow paths to the steam generators are maintained. The flow test will be conducted prior to entering Mode 3 so that the feedwater can be properly cleaned up before full power operation to minimize the impact on the steam generators. When performing tests required under 4.7.1.2 e.2, a dedicated individual'will be stationed at those manual valves, with direct communication to the control room, so that the system can be l restored to normal operable status if necessary.

Those valves that are locked in the AFW system are currently controlled by administrative procedures. With the proposed change, they will become a Technical Specification required item for verification. The changes as proposed, meet NRC staff's position as outlined in NRC letter dated February 21, 1984 (log No. 1455) The changes as proposed do not degrade the safety function of the AFW system.

Pursuant to the above evaluation, it is concluded that there is no unreviewed safety questions involved.

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Significant HazaEd Consideration This amendment request for the Surveillance requirement for the Auxiliary Feedwater (AFW) system does not represent a Significant Hazard. The proposed change includes valve position verification, 18 month flow verification, flow verification after modification or extended cold shutdown (>30 days in Mode 5) and an operator for valve realignment during

, testing.

The 18 month and extended outage (>30 days in Mode 5) flow path verifi-cation will provide adequate assurance'that the AFW flow paths to the Steam Generators (SG) are maintained. Flow path verification after maintenance and/or modification ensures that the AFW System is capable of delivering water to the SG.

When the plant is in Modes 1, 2 or 3 which requires Local Manual realign -

ment of valves that make the system inoperable, a dedicated individual shall be stationed at the valves (in communication with the control room) able to restore the valves to the normal operating status. _This ensures that the AFW system will be realigned should the safety system be required to function.

l The Commission has provided guidance concerning.the application of the standards in 10 CFR 50.92 by providing certain examples (48 FR 14870).

One of the examples of actions involving no significant hazards considera-tions related to a change that constitutes an additional limitation, restriction, or control not presently included in the technical specifica-i tions: for example, a more stringent surveillance requirement. (example ii)

The amendment request is in compliance with a request from the NRC dated February 21, 1984 (Log No.- 1455). These Technical Specifications are new

, restrictions and requirements that are not presently included in Davis-Besse Technical Specifications. These requirements ensure a flow path is t available from the water source through the AFW pumps to the S.G.'s and if required during certain testing the system will be realigned so as to perform its intended safety function.

Based on the above information, this amendment request would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

Therefore, based on the above, the requested license amendment does not present a Significant Hazard.

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J PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Two . independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

l ACTION:

a. With one auxiliary feedwater system inoperable, restore the .

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 ,

i 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 OPERABLE:

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

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

l 2. Verifying that each valve ( .n=1, power operated or

! automatic) in the flow path th;t i:; =t in'ad, :=1:d

.. in ==r:d tr. ;;;iti;r i r rao-Fke b.

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At least once per $8 months.y m r.; ;-... n , Sy:

PRIOR 70 ENTERING MooE a

1. Verifying that each automatic valve in the flow path actuates to its correct position on an auxiliary feed-water actuation test signal.

2. Verifying that each pump starts automatica11y upon receipt of an auxiliary feedwater actuation test signal.

3. VERIFylNG THATTHERE 15 A FLOW PATH BETWEEN KACH AUEILIARY FREDWATER PUMP AND EACM STEAM SENERATOR Sy PUMPING WATER FROM THE AUXI&lARY FEED PUMP TO THE STEAM SENERATOR. THE FLOW PATM 70THE STEAM SENERATOR JHALL BE WERIFIES Sy EITHER STEAM SENERATDR LEVEL CHANGE OR A0XILIARY FEEDWATER FLOW INDICATION.

, DAVIS-BESSE, UNIT VERIFlLATJON 1 Of THE 3/4 7-4 AUMILEARY WEEDWATER KWSTEM'S FLOW CAPA

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. . . PLANT SYSTEP.5 .

SURVEILLANCERiOUIREHENTS(Continued)

c. The Auxiliary Feed Pump Turb'ine Steam Generator Level Control System shall be demonstrated OPERABLE by perfomance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months. ,

d. The Auxiliary Feed Pump Suction Pressure Interlocks, and Auxiliary Feed Pusp Turbine Inlut Steam Pressure Interlocks snall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months. ,

S. AFTER ANY MODIFICATION OR REPAIR 70 THE AUXILit.Ry FEEDWATER SYSTEM THAT COULD AFFECT THE 5557EM*1 CAPAEILITU Te DELIVER WATER To THE STEAM SENERATom, THE AFFECTED FLOW PATH KNALL BE

. DEMONGTRATED AVAILABLE FRIOR TO ENTERING MODE.8 AC FOLLOWS$

! -l. IF THE MODIFICATieN OR REPAIR 11 DeWNSTREAM OF THETEET FLOW LINE,TNE AUKILIAR$ FEED PUM9 $NALL PUMP WATER TD TNE STEAM

• GENERATOR.; AND TNK Fi.DRPATH AVAILABILITY NILL SK VERIFIED BW ~

STEAM GENERAT.6R LEVEL CHANGE OR AUXfLIARY FEEDWATER FLOW

• INDICATION. ,

R. lFTNE MoblF'CATEDN OR REPAIR is UPSTREAM OFTNE TEET FLOW LINKY TNE AUKILIAR$ YEED PUMP SHALL PUMP WATER THROUGH TNE AUKELIARM FEEDWATER EySTEM 70 THE TEST FLOW LINEi AND THE FLOW PATH AVAILAEILIT.Y .WILL EE VERiflED EV FLOW sNDICATlaN IN TNE TEST FLOW LINEp (IEE NOTE SELOW)

VERIFICATION OF THE AUX 1LIARY FEEDWATER SYSTEM *J FLOW CAPACITY IS NOT RESulRED. ,

f. FDLLOWINS EACH EXTENDED COLD ENUTDOWN ( lP 20 day 3 IN MODE 5),8Y *

l. VERIFJANG THAT THERE IS A FLOW PATH BETWEEN EACH AUKILIARV FEED-WATER PUMP AND EACM STEAM GENERATOR 81 FUMPING WATER FR OM THE AUKILIARM FEED FUMP TO THE STRAM SENLR ATOR. THK, FLOW PATN TO THE KTE AM SENERATaR JNALL BE VERIFIED Ey EITHER STEAM GENER ATOR LEVEL CHANGE OR AUXILIARY FKEDWATER. FLOW INDIC ATIO N.

VERIFIC ATION OF THE AUKILIARM FqEDWATER SW.ETEM*1 FLOW CAPACITW IS NOT REQUIRED.

NOTE : WHEN CONDUCTING TESTS OF THE AUXILIARW FEEDWATER SYSTEM IN MODES 19 E, AND 2 WHICH RESUIRE LOC.AL MANUAL REALIGNME.NT DF VALVE 3 THAT MAME THE SYSTEM INDPERAELE, A DEDICATED INDIVIDUAL 5 HALT. SE ETATloNED AT THE VALVES (IN COMMUNICATION  ;

• WITN THE CONTROL ROOM) AELd TO RESTORE THE YALVES To NORMAL S3 STEM OPERAELE STATUS. Amendment No. 47, 63 DAVIS-BESSE, UNIT 1 3/4 7-5