Text

Forwards Application for Amend to License NPF-3 Revising Tech Specs to Reflect Changes Re Interim Anticipatory Reactor Trip Sys & Auxiliary Feedwater Flow Indications. Class IV Amend Fee Encl
	ML19225B246
Person / Time
Site:	Davis Besse
Issue date:	07/13/1979
From:	Roe L; TOLEDO EDISON CO.
To:	Reid R, Reid R; Office of Nuclear Reactor Regulation
References
	TASK-2.K.2.10, TASK-TM 527, TAC-45184, NUDOCS 7907240351
	Download: ML19225B246 (31)
	v • d • e

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txt iottoo Docket No. 50-346 G EDISON License No. NPF-3 LOWELL E. RDE Serial No. 527 v4. oreve.nz r .c i.t ... o,..w.-, t July 13, 1979 mei m sm Director of Nuclear Reactor Regulation Attention: Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Corcaission Washington, D. C. 20555

Dear Mr. Reid:

Under separate cover, we are transmitting three (3) original and forty (40) confomed copies of an application for Amendment to Facility Operating License No. NPF-3 for the Davis Besse Nuclear Power Station Unit No. 1.

This application requests that the Davis-Besse Nuclear Power Station Unit 1 Technical Specification, Appendix A,be revised to reflect the changes attached.

The proposed changes include 1) addition of requirenents on the interim antici-patory reactor trip system in sections 3.3.2.3, 4.3.2.3 and Table 3.3-15 and Table 4.3-15; 2) addition of requirements for auxiliary feedwater flow indica-tions in section 4.7.1.2; 3) change to react.or coolant high pressure reactor trip setpoint in Table 2.2-1; 4) addition of a reactor coolant pressure temperature curve, Figure 3.4-5; 5) addition of requirements in sections 3.4.3 and 4.4.3 for the electromatic relief valve; 6) change to technical specification basis 2.2.1; and 7) changt to SFAS actuatica mode and actions of Tables 3.3-3 and 4.3-2.

This amendment request involves several changes of Class III type. It is thore-fore determined to be a Class IV amendnent. Enclosed is $12,300.00 as requiced by 10CFR170.

The seven attachnents identify each propesed change, its safety evaluation and schedile requ; red to implenent the change after NRC approval. Items 1-5 above fulfill the seven day require,cnts of your letter of July 6, 1979.

Yours 3ry truly, LER:TJM Attachments THE TOLEDO EO:5CN COMPANY EOISCN PLAZA 300 MAO! SON AVENUE TOLEQQ. CH O 43ES2

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APPLICATION FOR AMENDPENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Technical Specifications, Appendix A to Facility Operating License No. NPF-3, together with the Safety Evaluation for the requested change. The proposed changes include

1) addition of requirements on the interim anticipatory reactor trip system in sections 3. 3. 2. 3, 4. 3. 2. 3 and Table 3.3-15 ard Table 4.3-15; 2) addition of requirements for auxiliary feedwater flow indications in section 4. 7.1.2 ;

3) change to reactor coolant high pressure reactor trip setpoint in Table 2.2-1;

4) addition of a reactor coolant pressure temperature curve, Figure 3.4-5;

5) addition of requirements in sections 3.4.3 and 4.4.3 for the electromatic relief valve; 6) change to technical specification basis 2.2.1; and 7) change to SFAS actuation mode and actions of Tables 3.3-3 and 4.3-2.

By ~/w ,M Vice I> resident, Faciliti'es Development Sworn to and subscribed before me this thirteenth day of July, 1979.

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e APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR F0WER STATION UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Technical Specifications, Appen!.ix A to Facility Operating License No. NPF-3, together with the Safety Evaluation for the requested change. The proposed changes include

1) addition of requirements en the interim anticipatory reactor trip system in sections 3. 3. 2. 3, 4. 3. 2. 3 and Table 3.3-15 and Table 4.3-15; 2) addition of requirements # auxiliary feedwater flow indications in section 4.7.1.2;

3) change to reamt coolant high pressure reactor trip setpoint in Tabic 2.2-1;

4) addition of a 1 m tor coolant pressure temperature curve, Figure 3.4-5:

5) addition of requirements in sections 3.4.3 and 4.4.3 for the electromatic relief valve; 6) change to technical specification basis 2.2.1; and i) change to SFAS actuati'n mode and actions of Tables 3.3-3 and 4.3-2.

By s/ Lowell E. Roe

~ lice President, Facilities Development Sworn to and subscribed before me this thirteenth day of July,1979.

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Notary Public LINDA L. CCCTELI.

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368 140

Docket "o. 50-346 License No. NPF-3 Serial No. 527 July 13, 1979 I Addition to Davis-Besse Nuclear Power Station Unit 50. 1 Technical Specif1-cations, Appendix A of Technical Specification 3. 3.2. 3, Table 3. 3-15 and Table 4. 3-15 conce rning the interin anticipatory reactor trip system. See proposed additions attached.

A. Time Required to Implement-This change can be effective upon NRC issuance.

B. Reason for Change ( Facility Change Request 79-283)

Th is is to identify limiting conditions for operation and surveillance requirements for thc new interin anticipatory reactor trip system.

C. Safety Evaluation This FCR calls for providing new technical specifications for

" Limit ing Condition for Operation" and surveillance requirements for the newly installed interim Anticipatory Reactor Trip System (ARTS) .

The new technical specifications are considered adequate to demonstrate and ensure the continued operability of ARTS. The sur-veillance requirements specified for this system ensure that the overall system functional capability is maintained and the periodic surveillance tests performed at the minimum frequencies are suf ficient to demonstrate this capability.

This is not an unreviewed safety question.

368 14i

TABLE 4.3-15 ANTICIPATOI Y REACTOlt TRIP SYSTill INSTI1U :ENTATION SUl:VEII. LANCE REOUIRI2'ENTS FUNCTIONAL UNIT CllANNEL MODES IN WilICll CHANNEL CllANNEL FUNCTIONAL SURVEILLANCE IS CllECK CALIBRATION TEST REQUIRED

1. Turbine Trip N.A. N.A. S/U* N.A.

2. Steam Generator - Feedwater S R 11 1,2 ,

Dif ferential Pressure - liigh **

3. Ourput Relay ' N.A. N.A. H 1,2 ,

This Surveillance will 1;e perforried prior to each unit start up if it has not been performed within the last 31 days.

This Surveillance requirencat is satisfied through surveillance requirerent 4.3.2.2.1.

368 142

s TAELE 3.3-15

. ANTICIPATORY REACTOR TRIP SYSTEM INSTRU'fENTATION FUNCTIONAL TOTAL NO. CHA'3ELS MINUTM APPLIC-UNIT OF 0F CHANNELS AELE CHANNELS TRIP OPERAELE MODES ACTION

1. Turbine Trip 1 1 1 1* 16

2. Steam Generator 1 1 1 1,2 17 Feedwater - Differential Pressure - High

3. Output Relay 4 2 3 1,2 18 ACTION 16 - With the number of channels OPERABLE one less than required by the Minirun Channels OPERABLE requirements, restore the inoperable channel to OPERAELE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce reactor power to less than 15% full power within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 17 - Uith the nunber of channels OPERABLE one less than required by the Minitun Channels OPERAELE requirements, restore the inoperable channel to OPE 2AILE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 18 - With the nucher of CPERA3LE channels one less than the Total Nurber of Channels, STARTUP and POWER OPERATION may procesd provided both of the following conditions are satisfied:

a) The control rod drive trip breaker associated with the inoperabic channel is placed in the tripped condition within one hour.

b) The Minitum Channels OPERA 3tE require:ent is cet; however, one additional control rod drive trip breaker associated with another channel may be tripped for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.3, after reclosing the control rod drive trip breaker opened in a) above.

Applicable only above 15% reactor power.

368 143

INSTRDENTATION ANTICIPATORY REACTOR TRIP SYSTEM INSTRDENTATION LI'f1 TING CONDITION FOR OPERATION 3.3.2.3 The Anticipatory Reactor Trip System instru=entation channels of Tcble 3.3-15 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-15 SURVEILLA'!CE REOUIRDENTS 4.3.2.3 The Anticipatory Reactor Trip Systen shall be deconstrated OPERAELE by the performance of the CILUC;EL C11ECK, CHANNEL CALIBRATION and CILUCIEL FUNCTIONAL TEST during the codes and at the frequencies shcun in Table 4.3-15.

368 144

II Changes to Davis-Besse Nuclear Power Station Unit Nc. 1 Technical Specifi-cations - Appendix A, Section 4.7.1.2 - concerning auxiliary feedwater flow measurement. See proposed changes attached A. Time Required to Implement This change can be ef fective upon NRC issuance B. Deason for Change (Facility Change Request 79-281)

This is to identify surveillance requirements for the new atixiliary feedwater flow indication.

C. Safety Evaluation The proposed surveilliance on this instrumentation is a 31 day CHANNEL FUNCTIONAL TEST and an 18 month CHANNEL CALIBRATION as defined in Davis-Besse Unit 1 Technical Specifications. This surveillance f requency is considered adequate to demonstrate and ensure the continued operability of the subject flow ins t r ument at ion .

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368 145

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERAT:f. _

3.7.1.2 Two inde - ;eam generator auxiliary feedwater pumps and associated flow paun shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one auxiliary feedwater system inoperable, restore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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.

2. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct cosition.

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b. At least once per 18 months, during shutdown, by:

1. Verifying that each aut.omatic valve in the flow path actuates to its correct position on an auxiliary feed-

. water actuation test signal.

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

. 3. ?e<44miv a C HANMEL CALIM ATioN on 4ke 0.utili3th icelwder bowmJtronea4dion ,

DAVIS-BESSE, UNIT 1 3/4 7-4

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Docket No. 50-346 License No. NPF-3 Serial No. 527 July 13, 1979 Page One of Two III Changes to Davis-Besse Nuclear Power Station, Unit 1 Technical Specifica-tions. Appendix A, Table 2.2-1 concerning the reactor coolant high pressure reactor protection system trip setpoint.

A. Time Required to Implement This change can be effective upcn NRC issuance

3. Reason for Chante (Facility Change Request 79-170)

The reactor coolant high pressure reactor protection systen trip setpoint was reduced in order to decrease the number of transients that could open the pressurizer pilot operated relief valve. This change reduces the upper bound of the current technical specification to be consistent with the recently revised setpoint.

C. Safety Evaluation This change provides for lowering the Reactor Coolant System (RCST high pressure trip setpoint f rom 2 351.4 psig to 2296.4 psig in the Re tor Protection System (PPS).

From the calculations performed by B5W it is evident that if the RPS high pressure trip setpoint is changed from 2351.4 psig to 2296.4 psig the peak pressurizer pressure cill remain below 2345 psig during loss of feedwater transients. The severity of these transients is reduced by implementing the change proposed by this FCR as described below.

In the case of loss of main feedwater the Steam and Feedwater Rupture Control System (SFRCS) isolates both steam generators (SG) on the feed-water and steam side, and auxiliary feedwater is initiated to the SGs.

Because cf the insufficient heat transfer to the secondary side of the SCs caused by loss of feedwater, the reactor is tripped on high RCS pressure.

It is estimated that with the present RPS high pressure trip setpoint the reactor will trip approximately 20 seconds af ter a loss of feedwater f rom 100% power as compared to 13 seconds with the lower RPS high pressure trip setpoint. Therefore, with the lower setpoint there is less energy (corresponding to the seven second time difference) contained in the RCS.

On the secondary side the steam pressure builds up because of main steam line isolation and the code safety valves lift (on the secondary side) to relieve the pressure. Since the energy content of the RCS is lower for the lower RPS trip setpoint, the secondary side code safety valves lift for a shorter duration. Consequently, when the secondary system code safety valves reseat, the secondary side level in the steam generators is higher for the lower RPS trip setpoint as compared to that which would be obtained with the present (higher) RPS trip setpoint. Also, when the level in the steam generator is brought to normal through auxiliary feedwater operation for the case of the lower RPS trip setpoint, a smaller amount of cold auxi-liary feedwater will be added to the steam generators, thereby reducing the net volumetric contraction of the RCS.

368 147

Docket No.30-346 License No. NPF-3 Serial No. 527 July 13, 1979 Page Two of Two C. Safety Eval.uation (Continued)

This will result in a higher minimum pressurizer level during such a transient with the lower RPS trip setpoint.

It should be noted that the RCS temperature and pressure remain the same af ter the secondary system code safety valves have reseated regardless of the higher or lower RPS high pressure trip setpoint.

This assumes that the pressurizer electromatic relief valve does not actuate during the transient. With the increase in the relief setpoint of the electromatic relief valve to 2400 psig, actuation of the relief valve will be eliminated during a loss of feedwater transient. Also, with the increase in electromatic relief valve actuation setpoint the amount of reactor coolant lost to the pressurizer quench tank on lif ting of the electromatic relief valve will be eliminated, leading to higher reactor coolant inventory in the RCS. to maintain a higher pressurizer level.

In su= nary, the proposed reduction in the RPS high pressure trip setpoint A s not degrade the safety of the plant and does not invalidate any of the safety analyses presented in the Davis-Besse Unit 1 FS AR or in the safety evaluation submitted to the NRC on Decenber 22, 1918 (Serial No. 475). The possaibility of an accident or a malfunction of a different type than any evaluated in the FSAR is not created. Also, the margin of safety as defined in the bases for technical specification is not reduced.

Pursuant to the above, the proposed change does not involve an unreviewed safety question.

Because 2300 psig in Table 2.2-1 is less than existing trip setpoint in license of 2355 psig, the setpoint change can be made prior to receiving the license amendment.

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Docket No. 50-346 License No. NPF-3 Serial No. 527 July 13, 1979 IV Change to Davis-Beasi Nuclear Power Station, Unit 1 Technical Specifications, Appendix A - Section 3.!'.9. 3, Figure 3.4-5 and Bases section 3/4 4.9 - conce rn-ing reactor coolant pressure and temperature limitations during emergency conditions - see attached proposed changes.

A. Time Required to In~1erent Diis change can be effective upon NRC issuance.

B. Reason for Change (Facility Change Request 79-287)

This change is proposed as a result of discussions with staff members of Nuclear Reactor Regulation and Office of Inspection and Enforcement.

C. Safety Evaluation See Attached 368 150

. ., Docket No. 50-346

, License No. NPF-3 Serial No. 527 July 13, I?79 Page One of Two SAFETY EVALUATION In certain small loss of coolant accidents, there may be no forced or natural circulatien of the reactor coolant, and water will be added to the reactor coolan:

sys:e by high pressure injectica pumps and/or by the makeup pumps. Under these energency/ faulted conditions, care must be taken to assure that pressure and temperature conditions at the reactor pressure vessel are maintained to avoid conditiens which could lead to propagation of flaws by brittle fracture.

During the situations of concern, it is assumed that normal RCS loop circula-tion may be interrupted. Under these conditiens, the RCS locp temperature sensors in the hot and cold legs cannot be relied upcn as accurate indications of tempera-ture ccaditions at the reactor vessel wall. Specifically, with interrupted flow, the cold leg reactor coolant systen temperature indicator, which is the normal point of reference for management of brittle fracture limits, cannot be relied upcn to reflect the reactor vessel conditions accurately. This is due to the fact that the ECCS injection point,.is down-stream of the RTD. Mcwever, it can be chewn for all conditicas in which tne reactor vessel is filled wita wa =r at 2-ast t- the level of the inlet no: les that the expected mixed avs ?e :e pera:ure in th a rocc or vcosel downcerer will be net less taan 1500F co'.d :r th : the tempera:uce at the core exit. This is true because the outle: plenum vent valves in Davis-Besse Uni: I will open in response to development of pressure differences resulting frca thermal temperature differences between the reactor vessel outle:

and inlet and permit the recirculation of steam or water through the ven valves and mixing with the cold water being injected frca the high pressure injec:icn systen prior to entering the downcocer. Given the 150 F temperature limit, :he operator may read an appropriate sampling of core exit thermocouples and inf er frca his measurement that downcomer temperature will be not more than 1500F colder than the temperature indicated by the core exit theraccouples.

L'ith the information available to hin, the operator should take a::ica if required, by reducing the injection rate of high pressure injection aater and/or takeup water to avoid exceeding appropriate pressure and temperature limits.

Appendix G of 10CFR50 states that this Appendix is "to provide adequa:e targins of safety during any condition of normal operation, including anticipated opera-ticnal occurrences and systen. hydrostatic tests,. . . ." Consequently, under the emer;cacy/ faulted ccnditions described above, the normal pressure-tempera:ure *-

li:1:s _o act appij. The modificc pressure-tempera:ure curve ic Figure 3.,-5 Lf7 was prepared to reflect the margins allowed under f aulted ccnditions. The curve *--

in Figure 3.4-5 still ccatains adequate margins to the real lini:s because a number of conservatists were used in the calculation. The use of a :cdified pressure-temperature curve is in accordance with paragraph 2.2.5 page' 5.3.2-i7 C*3

. of the NRC Standard Review Plan. The pressure and temperature limi:s in Fi;ure ND 3.4-5 were derised by removing the factor of two conservatis frc allowable 37) pressure which is prescribed by Appendix C. In all other respects, the curre has been calculated to meet the prescription of Appendix C for the Davis-Sesse Uni: I reactor vessel.

The pressure _ perature line for saturated water has been pict:ed in Figure 3.4-5 to indicate that this line is well below the acceptable limit curve for avoidance of reac:or vessel brittle fracture concerns during small LCCA events. Because the saturatica line is well away frc the brit:Ae fracture limit, it is evident that si;nificantly subcooled pressure and temperature conditiens , i.e. ,77 307, must be reached at the core outlet before the brittle fracture limits are approrched.

Having established these conditions, the operator may thrct:le high pressure in-jectica to =aintain the reactor coolant system pressure below the allowable pressure tc=perature curve and still be confident that adequate core cooling is being main-tained.

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Docket No. 50-346 License No. NPF-3 Serial No. 527

- -July 13,,19 79

', Page Two of Two The change does not constitute an unreviewed safety question because:

1. ~he probability of occurrence or the consequences of an accident or naltunction et equipnent inportant to safety, previousiv ~

evaluated in FSAR, has not been increased.

2. The possibility of an accident or nalfunction of a different t~e other than any evaluated previously in the FSAR has not been created.

3. The nagin of safety as defined in the basis for any technical spec.1 cation has not been reduced.

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REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTD1 D!ERGENCY/ FAULTED OPERATION

_ LIMITING CONDITION FOR OPERATION 3.4.9.3 In the emergency / faulted condition that there is no forced or natural circulation in the reactor coolant system and there is high pressure injection and/or makeup addition, the Reactor Coolant System temperature and pressure shall be limited in accordance with the limit line shown on Figure 3.4-5.

Under the above emergen..v/f aulted conditions, Figure 3.4-3 will not apply.

APPLICABILITY: Modes 3, 4, aad 5 ACTION: With the above limit exceeded, throttle the high pressure injection flow and/or makeup flow so that the pressure and temperature are within the acceptable limits within 30 mirutes.

SURVEILLANCE REOUIREMENTS 4.4.9.3 The Reactor Coolant System temperature and pressure shall be de t ermined o be within the acceptable limits at least once per 30 minutes durin:; the emergency condition.

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REACTOR COOLANT SYSTEM BASES The unirradiated transverse impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available.

The unirradiated impact properties and residual elements of the beltline region materials are listed in Bases Table 4-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced ART and the unirradiated RT The predicted ART are calculated usiktherespectiveneutronfkbbn.ceandcopperandpbhIphoruscontents.

Bases Figure 4-1 illustrates the calculated peak neutron fluence, at several locations through the reactor vessel beltline region wall and at the center of the surveillance capsules as a function of exposure time.

Bases Figure 4-2 illustrates the design curves for predicting the radiat. ion-induced ART as a function of the material's copper and phosphorus content an$Dkeutron fluence. The adjusted RT g 's of the beltline region materials at the end of the fifth full pokkr year are listed in Bases Table 4-1. The adjusted RT 's are given for the 1/4T and 3/4T (T is wall thickness) vessel wall khdations. The assumed RT of the closure head region is 40 F and the outlet nozzle steel forginhkT is 60 F.

During cooldown at the higher temperatures, the limits are imposed by thermal and loading cycles on the steam generator tubes. These limits are- segments D-E and D-F of the limit lines on Figures 3.4-2 and 3.4-4, respectively. These limits will not require adjustments due to the neutron fluences.

.; Figure 3.4-2 presents the pressure-temperature limit curve for normal heatup. This figure also presents the core criticality limits as

.- required by Appendix G to 10 CFR 50. Figure 3.4-3 presents the pressure-

- temperature limit curve for normal cooldown. Figure 3.4-4 presents the pressure-temperature limit curves for heatup and cooldown for inservice leak and hydrostatic testing.

The in Figues 3.9 -2. , 3 A- 3 ed 3 4 -4

-AM pressure-temperature limit curves,are applicable up to the fifth effective full power year. The protecticn against non-ductile failure is assured by maintaining the coolant pressure below the upper limits of Figures 3.4-2, 3.4-3 and 3.4-4.

368 155 1 DAVIS-BESSE, UNIT 1 B 3/4 4-11 e

REACTOR COOLANT SYSTEM EASES The limitations to prevent non-ductile f ailure during emergency / faulted operation when there is no forced or natural circulatica in the reactor coolant

. system and there is high pressure injection and/or makeup additien established.

Flyc. a.4-5 takcs into consideration that the reactor coolant system loop temperature sensors in the hot and cold legs cannot be relied upon as accurate indicaticas of temperature conditions at the reactor vessel wall. The pressure / temperature limi-taticas for this transient are given in Figure 3.4-5. The temperature scale in the curve has been shif ted upward by 150oF to conservatively account for the tempera-ture difference between the expected mixed average temperature in the reactor vessel downcomer and the temperature at the core exit. The pressure / temperature limits in Figure 3.4-5 were derived by removing the f actor of, two censervatism from allowable pressure which is prescribed in Appendix G. In other respects, the curve has been calculated *o meet the prescription of Appendix G.

n

. . i 368 156 3 3/4 4-11a

Docket No. 50-346 License No. NPF-3 Serial No. 527 July 13, 1979 V Change to Davis-Besse Nuclear Power Station, Unit 1 Technical Specifi-cations, Appendix A - Section 3.4. 3; 4. 4. 3 concerning pressurizer, electro-matic relief valve setpoints and the associated Bases section - see proposed change attached.

A. Time Required to Implement This change can be effective upon NRC issuance.

B. Reason for change (Facility Change Request 79-282)

This change adds the associated setpoints for the pressurizer electromatic relief valve.

C. Safety Evaluation See attached 368 157

Docket No. 50-346

. License No. NP F- 3 Serial No. 527 July 13, 1979 SAFETY EVALUATION For a RPS high pressure trip setpoint of 2300 psig, the na:<imu ovarshoot of the Reactor Coolant System pressure for a loss of feedaater (LOFL') event would be to 2350 psig. Also, the LOFW in the maximum over-pressure anticipa-ted transient. The string inaccuracies and drif t for the RPS high pressure trip are 15.29 psi, or 16 psi conservatively.

The inaccuracies and drift for the string that controls the electrc=atic relief valve for the pressurizer are 16.75 psi, or 17 psi conser- -

vatively. Included in this value is an inaccuracy of 4 psi and a drif t of 7.5 psi for the transmitter. The 4 psi and 7.5 psi were combined by takin; the square root of the sum of the squares, giving 8.5 psi. Subtracting 4 psi frca 8.5 psi gives a value of 4.5 psi that is attributable to only the drif t. "te 8.5 psi was then added to inaccuracy and drift values for other coepenents in the string to obtain a total of 16.75 psi.

The a}lowable value of h 2385._5 psig is obtained by subtracting 4.5 psi due to the drift from the trip setpoint ofh2390 psig. The minimum lift pressure for the pressurizer electro =ati: relief valve is then (2400-10-17) psi; = 2373 psig. Consequently, the resultant margin between the maximum pressure peak of 2366 psig and minimum lift pressure of 2373 psig for the pressurizer electromatic relief valve following an anticipated transient is 7 psi.

Tae above valuas for the pressurizer electromatic relief valve in conjunctica with a 2300 psig RPS high pressure trip setpoint will avoid actuation of the pressurizer electromatic relief valve during anticipated transients. All safety analyses for Davis-Besse Unit 1 assume that the vant capacity of th2 p res suri: 2r electromatic relief valve will not be available; thus, these analyses are unchanged by an increase in its setpoint.

The change dass not constitute an unreviewed safetv questien because:

. 1. The probability of occurrence or the consequences of an accident or calfunction of equipment important to safety, previously evaluated in FS AR, has not been increased.

2. The possibility of an accident' or malfunction of a different type other than any evaluated previously in the FSAR has not been created.

~

. 3. The margin of safety as defined in the basis for any technical specification.has not been reduced.

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q9 368 158 a eg,7 eat g

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REACTOR COOLANT SYSTEM SAFETY VALVES ::r' : 3 ANb ELECT?oAATic KE. Lie; VALVE - OPGEATie LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2435 PSIG + 1%.* Whes s,f isolaked,tke pr<uorita eMa~*bIc feliaG vcke. sLAll Ac.ve, i tr.* p .e 47 . 4. ..f A~ c2 3 9 c Ts t G and o^ nilowable.

APPLICABILITY: MODES 1, 2 and 3.

of K.SMtt**

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS Fa< %e. pre ssovsk.ee cede chh values, hve ue. no ,

4.4.3 Aadditional Surveillance Requirements other than those required by Specification 4.0.5. '

Fo< tLe pesso<i2.ee e l ec i<.mc+ic ce(te4 sclue a eV<l cal.h< dk <AeeI< z Lo,Il 6e pu +cined every I tnk.

" ine litt setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

" A h-c.ble vc! ve. fu c % eI a l.bre b cle & .

368 159 DAVIS-BESSE, UNIT 1 3/4 4-4

RFACTCR COOLM;T SYSTEM

/

/ BASE 3 f

For a RPS high pressure trip setpoint of 2300 psig, the maximum overshoot of the Reactor Coolant System pressure for a loss of feedwater (LOFU) event f would be to 2350 psig. Also, the LOFW is the maximum over-pressure antici- ,

pated transient.

The string inaccuracies and drift for the RPS high pressure 6

e. rip are 15.29 psi, or 16 psi conse rvatively. The maximum pressure peak for an anticipated transient is then 2366 psig. f i

i The inaccuracies and drif t for the string that controls the '

clactromatic relief valve for the pressurizer are 16. 75 psi, or 17 psi !

cor. s e rva t ive ly. Included in this value is an inaccuracy of 4 psi anj a drift of 7.5 psi for the transmitter. l The 4 psi and 7.5 psi were I combined by taking the square roct of the sum of the squares, giving 8.5 psi.

Subtracting 4 psi from S.5 psi gives a value of 4.5 psi that is attributable to only the drift. The 8.5 psi was then added to inaccuracy and drif t values for other components in the string to obtain a total of 16.75 psi.

The allowable value of h 2385.5 psig is obtained by subtracting 4.5 psi !,

due to the drift from the trip setpoint of h 2390 psig. The ninimum lif t pressure for the pressurizer electromatic relief valve is then (2400 17) psig = 2373 psig. Consequently, the resultant margin barween the maximum pressure peak of 2366 psig and minimum lif t pressure of 237. ,ig for the pressurizer electromatic relief valve following an anticipated transient is 7 psi.

Thus, a 2300 psig RPS high pressure trip setpoint and the above values for the pressurizer electromatic relief valve will avoid actuation of the pressurizer electromatic relief valve during anticipated transients.

368 160 B 3/4 4-la

Docket No. 50-346 License ."o. NPF-3 Serial No. 327 July 13, 1979 VI Change to Davis-Besse Nuclear Power Station, Unit 1, Technical Specifica-tions, Appendix A - Bases Section 2.2.1 (page B2-6) concerning revision to the basis of reactor coolant high pressure reactor trip (subject of Attachment III) and the reactor coolant pressure-temperature trip setpoints - see proposed change attached.

A. Time to Implement This change can be ef fect!ve upon NRC issuance.

B. Reason for Change (Facility Change Request 79-174)

This change is to update the bases section of the technical specifications to be consistent with current plant conditions.

C. Safety Evaluation This change calls for making the following changes to the Davis-Besse Unit 1 Technical Specifications:

1) On page B2-6, Reactor Coolant System (RCS) pressure-temperature trip setpoint should be changed from (13.01 T oyt F -5973) psig to (16.25 Tout F

- 7873) psig. This change was made to Table 2.2-1 of the technical specifications through license amendment 11 (dated June 16, 1978) following the removal of burnable poison rod assemblies and orifice rod assemblies from the core.

2) Through attachment, the Reactor Protection System (RPS) high pressure trip setpoint is being changed from 2355 to 2300 psig.

Safety evaluations have already been performed concluding that these are not unreviewed safety questions.

fbh

LIMITING SAFETY SYSTEM SETTINGS 3ASES The AXIAL POWER IMBALANCE boundaries are established in order to orevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/f t limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.

3 RC Pressure - Low, High and Pressure Temoerature

^

LO (L. The High and Low trips are provided to limit the pressure range in g anich reactor operation is permitted.

During a slow reactivity insertion startup accident from low cower d or a slew reactivity insertion from high power, the RC Hich Pressure sottnint is roachod before tM Hiah Flux Trip Setpoint. The trip set-d coint for RC High PressureLhD has been established to maintain

ne system pressure below the safety limit, 2750 psig, for any design transient. The RC High Pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lo.;er than the set pressure for these valves, 2435 psig. The RC High Pressure trip also backs up the High Flux trip.

f Tho.3C Low Pressure,1985 psig, and RC Pressure-Temperaturt(g-y.--

E ?5 % psig, Trip Satpoints have been established to maintain :ne Sl!!' ratio [reater than or equal to 1.32 for those design accidents that result in a pressure reduction. It also prevents reactor operation at cressures below the valid range of DNB correlation limits, protecting against DNB.

Hicn Clux/ Number of Reactor Coolant Pumos On f

In conjunction with the Flux - a Flux-Flow trip the High Flux / Number of Reactor Coolant Pumps On trip prevents the minimum core DNBR frca decreasing below 1.32 by tripping the reactor due to the loss of reactor

colant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

RE PLACE THE MATH EMA TIC AL y U PRE SSION W l'I N ( l 6 2 5 7os F- 78 73)

\

DAVIS-BESSE, UNIT 1 B 2-6 368 162

, e

Docket No. 50-346 License No. NPF-3 Serial No. 527 July 13, 1979 VII Change to Davis-Besse Nuclear Power Station, Unit 1 Technical Specifi-cations - Appendix A Tables 3. 3-3 and 4. 3-2 conce rning SFAS incident level 5 actuation - see proposed change attached

n. Time Required to Implement This change could be effective af te.r regional verification of equipment modification. This would be expected during the first planned outage of suf ficient duration eight months after NRC technical specification approval.

B. Reason for Change (Facility Change Request 79-171)

This change is proposed to reduce the possibility of a false SFAS incident level 5 trip in appropriately transferring decay heat and containment spray suctions to the containment emergency sump f or recirculation.

C. Safety Evaluation See attached 368 163

s Docket No. 50-346 License No. NPF-3

& ,1 %. W _

, , July 13, 1979 Page One of Two SAFETY EVALUATICN At the present tire Technical Specification 3.3.2.1 (Safety Features Actuation System), Table 3.3-3 requires that an inoperable Sorated Water Storage Tank (SWST) Level-low Instrument Strio.g or an inoperable Incident Level 5, Containment Su=p Recirculation Output Logic, be placed in the tripoed condition within one hour. This subjects the unit to a false t ril of Centainment Su=p Recirculation (Incident Level 5) if an addi-tior.a1 active f ailure is postulated. This false trip would produce a mora serious safety censequence than a f ailure to trip, as it would tra1sfer the suction of both crains of decay heat (DH) and containment spr..y (CS) pumps to the dry Containment Emergency Sump.

This request for a Technical Specification change proposes that the actien for the BNSi Level-low Instrument String be changed f rc 9 to a new 15 (at: ached). The action for Incident Level 5, Contain= cat Sump Recirculation Output Logic, will also be changed from 10 to 15. Also, Mode 4 should be eliminated f rem the applicable modes for the Inciden .

Level 5, Contain=ent Sump Recirculation Output Logic, since the Instru-nen String'for the BWST Level-low is not required in Mode 4 and there is no Inciden: Level 5 Manual Trip. The change from Action 9 to the new Actica 15 will allow the BWST Level-low Ins:rument String to be bypassed instead of tripped if it becomes inoperable and the Incident Level 5, Containment Surp Recirculation Output Logic, to be blocked if it is inoperable.

If a 3NST Level-low Instrument String is bypassed, the other three redundant SNST Level-low Instrument Strings can cause a 2/3 trip of SFAS Incident Level 5, which will cause the automatic transf er of the EH and CS pumps suctions from the BWST to the Containment Emergency Su=p.

If an Incident Level 5, Containment Surp Recirculation Output Logic is blocked, then one train of DH and CS pumps will not be automatically transferred from the BWST to the Contain=ent Emergency Su=p en an Inciden:

Level 5 trip.

The operator will then have to manually transfer this train of D3 and =cr CS pumps when the redundant train is automatically transferred. The sf) followirg alares and indications in the Control Roo vould indicate to e--

the operators when to =ake this manual transfer:

l. Control Room indicating lights and status board would indicate CI) that the Containment Surp Recirculation Cutput Logic was 'f3 blocked. m

2. Annunciator and computer alar:s would come on when SFAS started the HPI, DH and CS pumps.

3. There would be a: 1 cast three out of four BWST level indica-tions in the Control Room.

4. Annunciator and computer alarms would come on when the unaf-facted train was automatically transferred to the Contain=en:

Su=p.

Docket No. 50-346 License No. NPF-3

_ _ _ _ _ _ _ . . . Serial t:o. 527 July 13, 1979 Page Two of Two

-p

5. The operator would then =anually transfer the affected train as follows:

a. Stop HPI, DH and CS pu=ps in the train not transferred.

b. Close 3WST outlet valve (DH7A or B)

c. Open containment emergency su=p outlet valve (DH9A or 3)

d. Restart DH and ".S pumps.

c.

, Place HPI and DH. pumps in " Piggy Back" mode if necessary.

This manual transfer would have to be cade about 23 minutes af ter the initial SFAS trip that started all HPI, DH and CS pumps. This cnange will i= prove the safety of the unit by reducing the possibility of a f alse trip of SFAS Incidcat Leve'l 5, which would i= properly transfer . a suction of the EPI, DH and CS pumps te a dry Centain=ent E ergency Su=p. It is not an unreviewed safety issue.

~

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C RDemeck/ML 7/aj?g 368 165

u

$ d 'cgl S pec W c d ton c s. s.~3 e.prein ou s \3 g og ded by Telele i

ON -.

fdb l<Hu of 3[Ul71 B TABLE 3.3-3 s 5

T SAFETY FEATURES ACTUATI0ft SYSTEM IflSTRUMEilTATI0tl E

M MINIMUf!

. TOTAL fl0. UtlITS UtilTS APPLICABLE

. c- FullCTI0flAL UtlIT OF UtlITS TO TRIP OPERABLE MODES ACTI0tl 5

1. INSTRUMENT STRIflGS

]

a. Containment Radiation - "W* 13 #

liigh 4 2 3 All '1003 0#

b. Contain:nent Presture -

liigh 4 2 3 1,2,3 9#

c. Containment Pressure -

liigh-High 4 2 3 1, 2, 3 9#

d. RCS Pressure - Low 4 2 3 1, 2, 3* 9#

e. RCS Pressure - Low-Low 4 2 3 1, 2, 3** 9#

f. BWST Level - Low 4 2 3 1,2,3 X 16

2. CUTPUT LOGIC 5 a. Incident Level #1:

Containment Isolation 2 1 2 All MODES 10

b. Incident Level #2:

liigh Pressure Injection and Starting Diesel Generators 2 1 2 1,2,3,4 10

c. Incident Level #3:

Low Pressure Injection 2 1 2 1,2,3,4 10

d. Incident Level #4:

Centainment Spray 2 1 2 1.2,3,4 10

e. Incident Level #5:

Containment Sump Reci rcula tion 2 1 2 1,2,3,[ Jd IE

TABLE 3.3-3 (Continued)

TABLE NOTATION Trip function may be bypassed in this MODE with RCS pressure below 1800 psig. Bypass shall be automatically removed when RCS pressure exceeds 1800 psig.

- Trip function may be byoassed in this MODE with RCS pressure below 600 psig. Bypass shall be automatically removed when RCS pressure exceeds 600 psig.

- - One must be in SFAS Channels #1 or #3, the other must be in Channels 42 or #4

The provisions of Specification 3.0.4 are not applicable.

ACTION S'ATrMENTS ACTION 9 - With the number of OPERABLE functional units one less than the Total Number of Units operation may proceed provided both of the following conditions are satisfied;

a. The inoperable functional unit is placed in the tripped condition within one hour.

b. The Minimum Units OPERABLE requirement is met- how-ever, one additional functional unit may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 10 - With any component in the Output Logic inoperable, trip the associated components within one hour or be in at least HOT STAND 3Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ar.d in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 11 - With the number of OPERABLE Units one less than the Total Number of Units, restore the inoperable functional unit to O!ZRABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDCY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN with-in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 12 - a. With less than the Mimimum Units OPERABLE and reactor coolant pressure > 413 psig, both Decay Heat Isola-tion Valves (DHil and DHl2) shall be verified closed.

b. With less than the Mininum Units OPERABLE and reactor 368 167 coo, ant oressure < 413 psig operation may continue-hcwever, the furctional unit shall be OPERABLE prior to increasing reactor coolant pressure above 413 psig.

. ACT1oM 15- ( see n e x + p a $ c.)

DAVIS-BESSE, UNIT 1 3/4 3-12

s' ACTION 15 - With the number of OPERABLE units one less than t'le total number of units, bypass the inoperable func: tonal unit or block the inoperable output logic; and res: ore the inoperable functional unit or output logic to OPERACLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least :?0T STANDBY within the next six ho es and in COLD SHUTDC'.iN within the following 30 houre. With the number of OPERADLE .nf a one less than the total number of unics, one additional f unctional unit may be bypassed or biccked for up to two hours for surveillance teating per Speci-fication 4. 3.2.1.1 The following combinations of two output logics may be blocked simultaneously for surveil-lance testing:

'. Channels 1 and 3

2. Channels e and 4 No other conbinatious may be blocked simultaneously.

dh d/5 f

368 168

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Dear Mr. Reid:

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