ML20076D641

Proposed Tech Specs Providing Heatup & Cooldown Curves Applicable to 16 Full Power Years,Changing Overpressure Protection Sys Setpoint & Enabling Temp to Reflect Analysis Results

ML20076D641
Person / Time
Site:	Beaver Valley
Issue date:	07/15/1991
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20076D631	List: ML20076D626 ML20076D634 ML20076D641
References
NUDOCS 9107300104
Download: ML20076D641 (83)

Text

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REACTOR CQQLh11T_EYfi1Eli l REACIQ1L 22L>stiT PUliP_EFARTMP LIMITIllG CO!1DITIoli FOR OPERATIO!1 -,n_ ,_n,_ 3.4.1.6 If both OPPS PORV's are not OPERABLE, an idle reactor coolant pump in a non-isolated loop shall not be started, unless: pressu water level in less than 60 percent ), 4+e-a m/ ri z e r 1. The acgual l (840 ft 2. The secondary water temperature of each steam generator is less than 25'r above each of the in-service RCS cold leg temperatures. AEfLICABILITY: When the temperature of one or more of the non-isolated loop cold legs is 1 the enable temperature setforth in Specification 3.4.9.3 g egual $ AN With the pressurizer water level greater thanA 60 percent or theI temperature of the steam generator in the loop associated with the reactor coolant pump being started greater than 25' above the coldI leg temperature of the other non-isolated loops, suspend the startup of the reactor coolant pump. oe ega al fi fiMRVEILLhl[CE REOUIREliERTS. c'did'_~____ 4.4.1.6.1 The pressurl er water volume -m-the secondary waterI temperature of the non-isolated steam generators shall be determined within ten minutes prior to starting a reactor coolant pump. The secondary water temperature is to be verified by direct measurement of the fluid temperature, or contact tettperature readings on the steam generator secondary, or blowdown piping after purging of stagnant water within the piping. BEAVER VALLEY - UNIT 1 3/4 4-4a Proposed

l itR!AL PROPERTY BA515 4 CONikOLLING MTERIAL: INTERNEDIATE SHELL FLATE 86607 2 ) Ri F ER 9.5 EFPY: 1/4T, 202'F gy 3/47, 176'T / I i CURVES APPLICABL FOR HEATUP RATES UP TO 60'F/NR FOR THE SERVICE PERIOD UP TO f.5 EFPY. EJLJ I [ ' E I II I I I I II II I I Leak Test 2250 Limit f l 3 l 2000 i i 1750 Hea'typ Rates OgF/Hr J 7 To 60 y 1500 " " / r J J 1250 Unacceptable J J l Operation r 1000 f Criticality L'imit Based I f en'Inservic; L 2

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I l WAfft!AL PROPERTY BAS!$ \\ CONT t!NG MATERIAL: INTERNEDIATE SHELL PLATE B 6607 2 N07 AFTER 9.5 EFPY: 1/47, 202'r RT l \\ 3/47, 176'F \\ CURVES APPLICAB'lt FOR C00LDOWN RATES UP TO 100'F/HR FOR THE SER UP TO 9,5 EFPY, \\ N EJtJ I L m I I I I I I I I I I I I J ant I I I I I I I I I I J l I II I I I I 2 I I I J r I I I z 1 50 I I i I I I .] I I I F Iumnu y 9 I II y Un4CCeptable II E operation H' l I l1000 1 I J k gg$gh g I I I II FI I I I II I I F I I I I I I I I I I I I I I I I I 'I I I I I

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION m 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE: a. Two power operated rollof valvos (PORVs) with a nominal trip setpoint of 5 M V7.*t,4++-psig, or l b. A reactor coolant system vont of 2 3.14 squaro inchos. APPLICABILITY: When the temperature of one or more of the non-isolated RCS cold legs is s an enable temperature of-a4a'r. l 3/Y ACTION: a. With one PORV inoperable, oither rostore the inoperablo PORV to OPERABLE status within 7 days or depressurizo and vont the RCS through a 3.14 square inch vent (s) within the next 12 hours; maintain the RCS in a vented condition until both PORVs have been rostored to OPERABLE status. Rofer to Technical Specification 3.4.1.6 for further limitations. b. With both PORV's inoperable, depressurize and vont the RCS through a 3,14 square inch vont(s) within 12 hours; maintain the RCS in a vented condition until both PORVs have been rostored to OPERABLE status. c. The provisions of Specification 3.0.4 are not applicable, SURVEILLANCE REQUIREMENT t l 4.4.9.3.1 Each PORV shall be demonstratod-OPEDABLE Bi's l BEAVER VALLEY - UNIT 1 3/4 4-27a Proposed l' .. ~.. .~-..

7 3/4.4 REACTOR CQOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS, (continued) of Appendix G by -either- (1) restricting the water level in the and thereby providing a volume for the primary coolant to 6m/" pre,ssuri_zerintoor' (2) by restricting starting of the RCps to when the expand secondary water temperature of each steam generator is less than 25'T above each of the RCS cold leg temperatures. is removed from the isolated loop stop valves (hot leg and cold power leg) to ensure that no reactivity addition to the core can occur while the loop is isolated due to inadvertent opening of the isolated loop stop valves. Isolated loop startup is limited to Modes 5 and 6 in accordance with the NRC SER on N-1 loop operation. Verification of the isolated loop boron concentration prior to opening the isolated loop stop valves provides a reassurance of the adequacy of the shutdown margin in the remainder of the system. Restoration of to the hot leg stcp valve allows opening this valve to complete power the recirculation flowpath in conjunction with the relief line bypassing the cold leg stop valve and ensures adequate mixing in the isolated loop. This enables the temperature and boron concentration of the isolated loop to be brought to equilibrium with the remainder of the system. Limiting the temperature differential between the isolated loop and the remainder of the syste,m prior to opening the cold leg stop valve prevents any significant reactivity effects due to cool water addition to the core. loep Startup of an idle 44 will inject cool water f rom the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 20*F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient. 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizar code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 345,000 lbs. per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to l-__-_-_ BEAVER VALLEY - UNIT 1 B 3/4 4-la

ATTACHMENT B Beaver Valley Power Station, Unit No. 1 Propoced Technical Specification Change 110 191 i Revision of technical Specification 3.4.9.1 and 3.4.9.3 HEATUP AND COOLDOWil CURVES A!1D OPPS SETPOINT i 4 A. DESCRIPTION OF AMENDME!1T REQUEST The proposed amendment would extend the applicability of the heatup and cooldown curves to 16 effective full power years (EFPY) and modify the overpressure protection system (OPPS) pressure setpoint and enable temperature to incorporate analysis results applicable to 16 EFPY. In

addition, Specification 3.4.1.6.1, Su rveil la nce Requirement 4.4.1.6.1, and Bases 3/4.4.1 have been revised by changing "or" to "and."

i B. BACKGROUllD The proposed changes replace the current heatup and cooldown curves applicable to 9.5

EFPY, with curves applicable to 16 FFPY.

The new curves were developed by Westinghouco in accordance with Attachment E, "Duquesne Light Company Beaver Valley Unit 1 Life Attainment Plan." The curves provided in this plan have been developed in accordance with Regulatory Guide 1.99, Revision 2. Changes to the OPPS pressure setpoint and enable temperature were provided by our request for Technical Specification Change 11 o.

176, Revision 1 (TAC 76889) and are applicable to 9.5 EFPY.

The new OPPS setpoint and enable temperature are applicable to 16 EFPY. The new OPPS pressure setpoint was selected from the 16 EPPY curve developed by Westinghouse in accordance with i Attachment F, " Beaver Valley Unit 1 Low Temperature Overpressure l Protection System (LTOPS) Setpoint Analysis at 16, 24, 32 and 48 EFPY." The new enable temperature was determined in accordance with the criteria provided in Standard Review Plan (SRP) Section 5.2.2 Branch Technical Position RSD 5-2. During the review of a previous technical specification change

request, it was identified that a change to Specification 3.4.1.6 should be incorporated to ensure the requirements of both conditions in the limiting condition for operation are satisfied.

C. JUSTIFICATION The heatup and cooldown curves (Figures 3.4-2 and 3.4-3) have l been updated to extend the applicable requirements to 16 EFPY. The curves applicable to 16 EFPY were developed in accordance l with the same methodology used to develop the curves presently used. This will extend the applicability of these curves beyond the schedule for removal of the next surveillance capsule at 15

EFPY,

~

Proposed Technical Specification Change No. 191 - Page 2 ~ Specification 3.4.9.3 has been updated by providing a new OPPS pressure setpoint and enable temperature. The new OPPS pressure actpoint of 436 psig was selected from Attachment F Table 2 at 16 EFPY and 85*F. This is the most conservative setpoint defined by the 16 EPPY curve provided on Attachment F Figure 3. This setpoint has been reduced by 4'F to 432'P to address a nonconservatism identified in the new steam generator tube plugging analysis. The new OPPS enable temperature of 314*F was determined in accordance with the relationship RT + 90*P l NDT provided in SRP section 5.2.2 Branch Technical Position RSD 5-2. As shown on the 16 EFPY heatup and cooldown curves.the limiting is at the 1/4t location and is 224*F, therefore, the new RTNDT enable temperature is 224

• F. +90
• F = 314
• F.

The intent.of-Specification 3.4.1.6 is to provide an alternate set-of conditions in the event a reactor coolant pump must be started when-both-PORV's are not operable. Specification j 3.4.1.6.1, surveillance Requirement 4.4.1.6.1, and Bases 3/4.4.1 have been revised from "or" to "and" to ensure both conditions are in effect to mitigate the consequences of starting a reactor coolant pump. The action statement of specification 3.4.1.6 has been revised to state " greater then or equal to" for both the pressurizer level and temperature differential to address the full range of setpoint conditions. D. SAFETY ANALYSIS Attachment E was developed by Westinghouse to provide additional heatup and cooldown curves for vessel exposures extending to 48 EFPY.- The current heatup and cooldown curves are applicable to 9.5. Plant operation will exceed 9.5 EPPY during Cycle 9, - theret-new curves applicable to 16 EPPY and determined in achou with the same methodology used to produce the 9.5 EFPY curv is

o developed by Westinghouse.

These curves will be used - to addr"es plant operation until new curves can be generated based-on thn examination-of the next capsulo removed. The new OPPS - pressure setpoint and enable temperature were determined using the same methodology used to determine the values provided in Technical Specification Change No. 176 Revision 1 -(TAC 76889). Attachment F was developed by Westinghouse to provide additional OPPS setpoints for vessel exposures extending to 48 EPPY.- These OPPS curves could-be used in-determining applicable-setpoints based on current plant pressure and temperature conditions, however, the plant does not installed system to a'tomatically perform this function. have _ an u l - Technical Specification Change No. 176 Revision 1 incorporated a - single pressure setpoint based on the lowest pressure on the 9.5 e . EFPY curve at 85*F. This setpoint has been updated by incorporating a single pressure setpoint based on the lowest pressure on the 16 EFPY curve at 85'F, as shown on Attachment F Table 2,. this is 436 psig. A nonconservatism was identified during the now steam generator tube plugging analysis performed by Westinghouse, therefore, to climinate this condition the setpoint has been reduced to 432 psig. .e Ye' su' Wv quyri -uire av pp=p=h 9= 7-weWr +T='-'W=='Md'"F-WW 'w'Drm -4=D% t'w1*y**MN "' w = 97 M Wwr M c-W '-"9gunTr*T'- f Fw ys-ng e T M' +- g'N^vm-T$' y M WeraMW 'T*-2x"'T'W7*P-45 TFb'( w V y

l Proposed Technical Specification Change 11 o. 191 Page 3 The OPPS enable temperature in based on the limiting reactor vencel material RT Au shown on the new heatup and cooldown gyp. in 224'F, this 10 different from the 202*r curves the RTk)j gyI.,, E F PY, therofore, the OPPS enable temperature han value on the been updated to incorporate a new value applicable to 16 ErPY. The consequences of a heat input transient caused by the addition when utarting a reactor of energy from the secondary system. lho OPPS. The reactor coolant pump can be mitigated without using coolant system will be protected againnt overpreunure tranniento and will not exceed the limits of 10 CrH 50 Appendix G by (1) restricting the precourizer water level to provide a volume for the primary coolant to expand

into, and (2) restricting the temperature differential between the reactor coolant cyctem and the steam generator to limit the addition of energy transferred to the reactor coolant nyatom.

Changing the Limiting Condition for Operation, the curveillance requirement and the basen from "or" to "and" ensuren the required conditionn are natintled in accordance with the accident analyces. The action ntatement conditions of specification 3.4.1.6 have been changed from " greater than" to " greater than or equal to." Specification 3.4.1.6 requiren pressurizer level less than 60 percent and temperature differential less than 25*P, therefore, the action statement han been revised to ensure the upecification addrenceu the iulI range of setpoint conditionn including these at, above and below the specified netpoint. Based on the above considerations, these changen reflect the application of methodologiou recognized by the 11HC and the industry as providing a sufficient margin of nafety. The fracture toughness requirements of 10 CPR 50 Appendix C are satisfied and conservative operating restrictionn are applied in the proposed heatup and cooldown

curves, therefore, these changes are considered to be safe and will not reduce the safety of the plant.

E. 110 SIG111 FICA 11T llAZArtDS EVALUAT10!1 The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards act forth in 30 CFR 50.92(c) as quoted below: The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a facility licenned under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequencen 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 nignificant reduction in a margin of cafety.

Proposed Technical Specification Change No. 191 Pago 4 Tho following evaluation is provided for the no significant hazards consideration standards. 1. Does the change involve a significant increase in the probability or consequencos of an accident previously evaluated? The heatup and cooldown curvos have boon revised to extend their applicability from 9.5 EPPY to 16 EFPY. The new hoatup and cooldown curvos woro developed in accordance with the methodology provided in Regulatory Guido 1.99 Revision 2. .This is consistent with the methodology used to determino the current curvos. The OPPS notpoint and enable temperature have also been updated to extend their applicability to 16 EPPY. An analysis was. performed by Westinghouse to provide additional OPPS sotpoints for vessel exposures extending to 48 EFPY. - Thoso OPPS curvos could be used in datormining applicable notpoints bauod on current plant pressure and temperaturo conditions,

however, the plant does not have an installed system to automatically perform this function.

Technical Specification. Change No. 176 Revision 1 incorporated a single pressure setpoint based on the lowest pressure on the 9.5 EFPY curve at 85'F. This sotpoint has been updated by incorporating a single pressure sotpoint based on the lowest pressure on the 16 EFPY curvo at 85'F, this results in a sotpoint oof 436 psig. A nonconservatism in the OPPS setpoint was identified during the new steam generator tubo pluggingi analysis, thereforo, Westinghouse has determined that reducing the sotpoir.t by 4*F to 432*F will eliminato this condition. The OPPS onable temperaturo in based on the limiting reactor vessel material RTHDT.. The now heatup and cooldown. curves ~, specify a RTNbT of 224*F versus 202*F on the current curvos. SectIon 5.2.2 of the Standard Review Plan includes Dranch Technical Position RSD 5-2 which providos guidance for datormining the OPPS onabla temperature. This mothod i was used:.and results-in an onable temperature of 224*F + 90*F = 314"F. i Tho-consequences 'o f a heat -input transient caused by the addition of energy from the secondary system when starting a reactor l coolant pump can be mitigated without using the g OPPS. -The reactte ccolant system will.be protected against overpressure-transients and will not exceed the limits-of 10 CFR 50 -Appondix G by (1) rostricting the prosaurizer wator-lovel to provido a volume for the primary coolant to expand' into,- and (2) restricting the temperature differential between _the reactor coolant system and the steam generator . to limit the addition of energy transferred to thc roactor coolant system. Changing the Limiting Condition for Operation, the. surveillance requirement and the Bases from "or" to "and" onsures the required conditions are satisfied in accordance with the accident analyses. The action -gp agtga w us-re.awnywm--+-e<-p,i-ggewWw, yay _wwww., p.y-wmhw,w_w w -

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• Proposed Technical Specification chango No.-191

) Page 5 statchiont conditions of Specification 3.4.1.6 have been changed from "greator than" to "greator than or equal to." Specification 3.4.1.6 requires pressurizer lovel less than 60 percent and temperature differential loss than 25*F, thorofore, the action statomont has been revised to ensure the specification addressos the full rango of setpoint conditions including those at, above and below the specified setpoint. These-changos woro determined in accordance with the methodolo set forth in the regulations to provido an adequato .tgin of safety to ensure the reactor vossal will withstand the offects of normal cyclic loads due to . tomperature and pressure. changes as-well as -the loads associated with postulated faulted events. Thereforo, the proposed changes will not significantly increase tho - probability or consequences of an accident previously evaluated. 2. - Doco the change create the possibility of a now or different kind of accident from any accident previously evaluated? -The new heatup and cooldown curves were developed in accordance with the methodology used to determine tho current c'arves and-are consistent with the methodology set forth -in-the regulations. The now OPPS pressure setpoint i was selected from the most conservative position on the 16 EFPY low temperaturo overpressure protection setpoint curve to-ensure sufficient margin is available-to-prevent violation of the pressure-temperature _ limits due to anticipated mass and heat input transients.- The new onable temperature provides a wider range over-which the OPPS is -active and was determined in accordance with the regulations. The intent of Specification 3.4.1.6 is to provide.an alternato' set of conditions in the event a reactor coolant pump must be started when both PORV's are_not operable. Specification' 3.4.1.6.1, Surveillance Requiremont 4.4.1.6.1, and_ Bases 3/4.4.1. have been revised from "or" to "and" to ensure both -conditions are-in offect to mitigate the consequences of starting a reactor _ coolant-pump. These changes are consistent with the regulations and will not affect the 'rollabil ~. ty. of the reactor vessel or'the plant heatup and cooldown procedures. Therefore, the - proposed ' changes will not create the possibility of a new or different - evaluated.. kind 'of accident from any accident previously-- - 3. Does the change involve-a significant reduction in a mar <jin of safety? i MNg ' ty -' i t-MYM -v'n-7M $ 1'" 79e -Ymw--- vr yeg-y+--rm e-iwmr*s 1 +N-y r3t-rer ..weRTyrw r -Tw=were w p-w qg -e eveM--etr'-- 9Mt+-h+s7Wrir r.,. T

Proposed Technical; Specification Change No. 191 Page 6 The revised hoatup and cooldown

curves, OPPS pressuro

setpoint, enable temperature, and change to Specification 3.4.1.6 will cer.tinuo to onsure the reactor coolant system will be protected from prossure transients at low temperaturos.

The proposed changer will not reduce the reliability of the

OPPS, nor will they increaso the likelihood of vossol damage or failure in the event of an ovorproscure transient.

Those changes are established in accordanco with current regulations and the latest regulatory guidanco. Plant oporation will be maintained within required

limits, thoroforo, the roactor vossol materials will behavo in a

non-brittle manner to remain within the plant-design basis. Thorofore, the proposed changes do not involvo a significant reduction in a margin of safety. F. NO SIGNIFICANT liAZARDS CONSIDERATION DETERMINATION L Based on the considerations expressed above, it is concluded that the activitics-associated with this licenso amendment request satisfies the no significant hazards consideration standards of 10 CFR 50.92(c) and, - accordingly, a no significant hazards consideration finding is justified.- G. ENVIRONMENTAL EVALUATION The proposed-changes have been ovaluated and it has boon determined.that the changes do not involvo (1) a significant hazards consideration, (ii) a significant change in the types or t significant-increase in the amounts of any effluents that may be-i_ released

offsito, or (iii) a significant incroaso in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes moot the eligibility criterion for catogorical exclusion set forth in 10 CFR 51.22(c)(9). Thorofore, pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed changes is not required. H. UFSAR CHANGES Reference to Attachment F has been added to UFSAR section 4.2. I. L l h E ,n \\.

ATTACllMI'llT C i Beaver Valley Power Station, Unit IJo, 1 Proposed Technical Specification Change 11o. 191 f Typed Pages: 3/4 4-4a 3/4 4-24 3/4 4-25 3/4 4-27a il 3/4 4-la 1 6 L 1

REACTOR COOLANT SYSIDj BEACTOR COOLANT PUMP STARTUt LIMITING CONDITION FOR OPERATION 3.4.1.6 If both OPPS PORV's are not

OPERABLE, an idle reactor coolant pump in a non-isolated loop shall not bo started, unless 1.

The acgual pressurizer water level is less than 60 percent (840 ft ), and l t 2. The secondary water temperaturo* of each steam generator is less than 25'F above cach of the in-sorvice RCS cold log i temperatures, j APPLICABILITY: _ When tne temperature of one or more_of the l non-isolated loop cold legs is 5 the enable temperature setforth in Specification 3.4.9.3. l ACIlRH: l with the pressurizer water level greater than or equal to 60 percent l or the temperature of the steam generator in the loop associated with the reac cor__ coolant pump being started greater than or. equal to 25* l above the cold leg temperature of the other non-isolated loops, suspend the startup of the reactor coolant pump. t SURVE1LLANCE REQUIREMENTS 4.4.1.6.1 The pressurizer water volume and the secondary water l temperature of the non-isolated steam generators shall be determined within ten minutes prior to starting a reactor coolant pump. r - The secondary -water temperature is - to be verified tar direct measurement of -the-fluid temperature,- or contact temperature readings on the steam generator secondary, or blowdown piping after purging of stagnant water within the piping. L 1 1 BEAVER VALLEY - UNIT 1 3/4-4-4a Proposed ._ w - ..,-m.d.,m., ...w,-A._-.'-..,---w- ,,.+..m

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MATERIAL PROPERTY BASIS CONTROLLING MATERIAL; INTERMEDIATE SHELL PLATE B6607 2 RTNDfFTER 16 EFPY; 1/4T,224 'F 3/4T,188 'F 2500 LEAK TEST 2250 ~ LIMIT 2000 ~ I n0 G5 Q1750 HEATUP RATES / TO 60 'F/HR UNACCEPTABLE [ CRITICALITY Q OPERATION / LIMIT BASED fogo / ON INSERVICE h 73, / HYDROSTATIC o TEST TEMP, E 329 'F FOR s ACCEPTABLE THE SERVICE OPERATION PERIOD UP

1. 50 TO 16 EFPY 0

0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (*F) FIGURE 3.4 2 Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 16 EFPY BEAVER VALLEY UNIT 1 3/4424 PROPOSED

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL; INTERMEDIATE SHELL PLATE B0607 2 RTNDAFTER 10 EFPY; 1/4T,224 'F 3/4T,188 'F 2500 2250 2000 n9 $1750 [ UNACCEPTABLE OPERATION cr a 1500 $1250 ct Q1000 COOLDOWN 75o y 5 RATES 'F/HR ACCEPTABLE E o% / OPERATION 500 60 ' s 250 100 ' o 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE ( F) FIGURE 3,4 3 Beaver Valley Unit 1 Reactor Cor!ent System Cooldown Limitations Applicable for the First 16 EFPY BEAVER VALLEY. UNIT 1 3/4 4 25 (next page is 3/4 4 27) PROPOSED

• REAC101LCQQ1 ART _S.Y11TE11 l

9.YXRPltEfiE2RE PRQTICILQ1LfullTflis I i L1141 TING CONDITION FOR OPERATIO!J n.4.9.3 At least one of the following overpresnure protection nystems shall be OPERABLE: power operated relief Valves (PORVs) with a nominal trip l a. Two setpoint of s 432 psig, or b. A reactor coolant system vent of 2, 3.14 square inches. AEELLCABILIIX: When the temperature of one or more of the. I non-isolated RCS cold legs is s an enable temperature of 314*P. l l AC31Gif1 a. With -one PORV inoperable, either restore the inoporable PORV to OPERABLE status within 7 days or depreusurize and vent the RCS through a 3.14 square inch vent (s) within the next 12-hours! -maintain the RCS in a vented condition until both PORVs have. been restored to OPERABLE status. Refer to Technical Specification 3.4.1.6 for further limitations. b. With both FORV's inoperable, depressurize and Vent the RCS through a 3.14 square inch vent (s) within 12 hours; maintain-the RCS in a vented condition until both PORVs have been restored to OPERABLE status, c. The provisions of specification 3.0.4 are not applicable. SURVEILLANCE REQUIREN ENT i 4.4.9.3.1 Each POPV shall be demonstrated OPERABLE BY: l' DEAVER VALLEY - UNIT 1 3/4 4-27a Proposed _. _., _,. _,. ~.. _

3/4.4 REACTOR COOLANT SYSTEli e' BASES nawa -- LM.1 REACTOILC2Q1 ANT LOOPS, (continued) of Appendix G by (1) rostricting the water level in the pressurizer l and thereby providing a volume for the primary coolant to_ expand into and (2) by restricting starting of the RCPs to when the secondary l water temperature of each steam generator is less than 25'F above each of_the RCS cold leg temperatures. Power is removed from the isolated loop stop valves (hot leg and cold leg) to ensure that no reactivity addition to the core can occur while the loop is isolated due to inadvertent opening of the isolated loop stop valves. Isolated loop startup is limited to Modes 5 and 6 in accordance with the NRC SER on N-1 loop operation. -Verification of-the isolated loop boron concentration prior to opening the isolated loop stop valves provides a reassurance of the adequacy of the shutdown margin in the remainder of the system. -Restoration of power to the hot leg stop valve allows opening this valve to complete the recirculation flowpath in-conjunction with the relief line bypassing the cold leg stop valve and ensures adequate mixing in the isolated. loop This enables the temperature and boron concentration of -the isolat,a loop to be brought _to equilibrium with the remainder of the system. Limiting the temperature differential between the isolated loop and the remainder of the system prior to opening the cold leg stop valve prevents any significant reactivity effects due to cool water addition to the core. Startup of an idle loop will inject cool water fron the loop into the core.- The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature 'is within 20'F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike-which could result from this cool water induced reactivity transient. 3/4,.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from .being pressurized above its' Safety Limit of 2735 psig. lEach-safety valve is designed to'rolieve 345,000 lbs. per hour of saturated-steam at-the valve set point, The relief capacity of a single safety valve is adequate'to BEAVER VALLEY - UNIT 1 B 3/4 4-la Proposed

ATTACHMENT D Beaver Valley Power Station, Unit No.1 Proposed Technical Specification Change No.191 I UFSAR CHANGES I I l l l l i l I 1 l l

i DVPS-1-UPDATED FSAR Rev. 2 (1/84) References for Section 4.2 1. Ernest L. Robinson, " Bursting Tests of Steam-Turbine Disk Wheels," Transactions of the ASME, (July, 1944). ) 2.- D. H. Winne, B. M. Wundt, " Application of the Griffith-Irwin Theory of Crack Propagation to the Bursting Behavior of ) Disks, Including Analytical and Experimental Studies", ASME i (December 1, 1957). 3. J. W.

Murdock,

" Performance Characteristics of Elbow Flowmeteres," Transactions of the ASME, (September, 1964). 4. J.-J. Szyslowski, R. Salvatori, " Determination-of Design l Pipe Breaks for Westinghouse Reactor Coolant Systems" WCAP-7503 Revision 1, Westinghouse Electric Corporation (February, 1972).-- 5. Duquese Light Company to NRC submittal concerning NUREG-0737, Item II.D.1 Pressuricer Safety and Relief Line Piping and Support Evaluation dated June 24, 1983. i 6. Duquosne-. Light Company to NT submittal concerning NUREG-0737, Item II.D.1 Plant Specific Report dated July 1, 1982. l l b ea ve g lla lft

• tefl

),0 0 Te + 14e"ob<rM hgq, ),1 s,5 u.9 ff9;k(;bn' sysh.c(L7vh) sed w& A ul & d n,2 g n a J n wh y [L t'~5 YI It n s 2. ),e 2 e,C DL tt) (~15", Jam,, y 2 3, / ff d, I (. L i W l 4,2-24 -.. = -.. a.-.- .... - - -. ~.... -. -. - - -

i j j i i ATTAL,,f MJNT E i Beaver Valley Power Station, Unit No.1 Proposed Technical Specification Change No.191 Duquesne Light Company Beaver Valley Unit 1 i Life Attainment Plan 1 i I l I i =_ i E _. _. _. - -. - - _., _.. _ _ -. ~. _. _. J

4' DUQUESNE LIGHT COMPANY BEAVER VALLEY UNIT 1 LIFE ATTAINHENT PLAN N. K. Ray J. H. Chicots February 1990 / Approved by: T. A. Heyer, Manager Structural Materials & Reliability Technology Hork Performed for Duquesne Light Company HESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 . Pittsburgh, Pennsylvania 15230-2728 o 1990 Westinghouse Electric Corp. 06520:10/063090 m m my -- w 'N+.- mg--g u y= u-F-e +r ywm-n -e m r-y1 --N*y-- y

TABLE Of CONTENTS But i Table of Contents 11 List of Tables ill List of Figures 1-1

2.0 DESCRIPTION

Of KEY ISSUES-2-1 3.0 IRRADIATION EMBRITTLEMENT PREDICTIONS 3-1 h 3.1 Pressurized Thermal Shock Hethodology 3-1 3-3 3.2 Regulatory Guide 1.99, Revision 2 4-1 4.0 PLANT SPECIFIC EVALUATION 4.1-Pressurized Thermal Shock Evaluation 4-1 4-6 4.2 Heatup and Cooldown Curves 5-1 5.0 FLUX REDUCTION GOAL 5-1 5.1 Fluen;e Limits. l l 5.1.1 Pressurized Thermal Shock 5-1 5.1.2 Heatup and Cooldown Curves 5-3 5.1.3 Emergency Response Guide Line Limits 5-5 6-1 6.0. CONCLUSION 7-1

7.0 REFERENCES

l l l 06520:10/062090 1

LIST OF FIGURES I h92 f_iSEE 2-1 Effect of Irradiation on.the Charpy V-notch Toughness 2-3 2-2 Pressure-Temperature Heatup and Cooldown Operational 2-4 Limit Curves (Sample) 4-1 Identification and Location of Beltline Region Material 4-2 the Beaver Valley Unit i Reactor Vessel. for Beltline Region Materials 4-4 4-2 Fluence Vs RTPTS (PTS Rule) For Beltline Region Materials 4-5 4-3 Fluence Vs RTPTS (Reg. Guide 1.99. Rev. 2) 4-8 44 Beaver Valley Unit i Vessel Coolant System Cooldown Limitations ApplicaMe for the First 9.5 EFPY, using Reg. Guide 1.99, Rev. 2 (Current Curves) (No margin for instrument error) 4-9 4-5 Beaver Valley Unit i Vessel Coolant System Heatup Limitations Applicabie for the First 9.5 EFPY, Using Reg. Guide 1.99, Rev. 2 (Current Curves) (No margin for instrument error) 4-10 4-6 Beaver Valley Unit i Vessel Coolant System Cooldown Limitations Applicable for the First 16 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-11 4-7 Beaver Valley Unit 1 Reactor Vessel Coolant System Heatup Limitations Applicable for the First 16 EFPY Using Reg. Guide 1.99, Rev. 2. ('No margin for instrument error) iii 0652D:10/062090 l

LIST OF FIGURES (continued) figEt EA9.e 1 4-8 Beaver Valley Unit i Vessel Coolant System Cooldown 4-12 Limitations Applicable for the First 24 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-9 Beaver Valley Unit ' Reactor Vessel Coolant System Heatup 4-13 Limitation; Applicable for the First 24 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) i 4-10 Beaver Valley Unit 1 Reactor Coolant System Cooldown 4-14 Limitations Applicable for the first 32 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-11 Beaver Valley Unit 1 Reactor Coolant System Heatup 4-15 Limitations Applicable for the First 32 EFFY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-12 Beaver Valley Unit 1 Reactor Coolant System Cooldown 4-16 Limitations Applicable for the First 48 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-13 Beaver Valley Unit i Reactor Coolant System Heatup 4-17 Limitations Applicable for the First 48 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 5-1 Required Flux Reduction Factor Using PTS rule 5-4 5-2 Acceptable LTOPS Set Point Range Vs. RCS Temperature 5-7 06520:10/062090 iv m____.

(.

1.0 INTRODUCTION

I Neutron embrittlement represents the most significant damage mechanism that could potentially limit the lifetime of the reactor vessel. This report will examine the various issues which are affected by irradiation damage, and examine the potential'_need for flux reductions to demonstrate reactor vessel l integrity for different time spans, including life extension of the Beaver Valley Unit i Reactor vessel. RT values were calculated using PTS rule [1] and Regulatory Guide 1.99, PTS Revision 2 [2] using material chemistry of Table 1-1. Table 4-1, indicates i that the lower plate 86903-1 is the most limiting material for the Pressurized Thermal shock evaluation. RT Vs fluence plots are developed using PTS PTS rule and Reg. guide 1.99, Rev. 2 methodology and are shown in Figures 4-2 and 4-3. These figures are applicable as long as (1) material chemistry remains unchanged and (2) the rule for calculation of PTS reniains the same. The calculation for RT for the beltline region of the Beaver Valley NDT Unit I reactor vessel indicates that the limiting material for developing ( heatup and cooldown curves, are the lower plate B6903-1 and intermediate plate B6607-2 for 16, 24, 32 and 48 Effective Full Power Years (EFPY). RTNDT l l1 values for the limiting material is reported in Table 4-2. values for The flux reduction goals were carried out to insure 1) the RTPTS the Beaver Valley Unit I reactor vessel materials remain below the screening criteria for thermal shock and 2) comfortable margin (identified by the Beaver Valley Unit I personnel) remans up to the end of 48 EFPY. l Table 4-1 indicates that the RT value for B6903-1 for 4B EFPY is above PTS the PTS screening criteria using PTS rule. For this, flux reduction is needed only.for the lower plate B6903-1 using PTS rule. Using Reg. Guide 1.99, Rev. 2, it has been shown in Table 4-1 that the PTS values for the Beaver Valley Unit I remain below the PTS screening criteria up to 48 EFPY. 06520:10/063090 1-1

Also, it has been revealed by the Duquesne Light personnel that by adjusting the low temperature overpressurization ('. TOP) system set point, they will have comfortable margin to operate tne reactor vessel up to the end of 48 EFPY, assuming 1) no increase in flux and 11) no significant increase in material chemistry of the belt line region. 1-2 06520:10/063090

TABLE 1-1 BEAVER VALLEY UNIT 1 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES I(a) g(b) Cu Ni (Ht.%) (Ht.%) CF (*F) (*F) Int. Plate, B6607-1 .14 .62 100.50 43 34 Int. Plate, 86607-2 .14 .62 100.50 73 34 Lower Plate, B6903-1 .20 .54 141.80 27 34 (167.9)c j7(d) Lower Plate, B7203-2 .14 .57 98.65 20 34 Long. Held, 305424 .28 .63 191.65 -56 65.5 (191.4)c 44.05(d) Long. Held, 305414 .34 .07 210.45 -56 65.5 Cire. Held, 90136 .29 .61 132.90 -56 65.5 NDT (I) values for the forgings are measured and initial (a) The initial RT RT value for the weld is generic. NOT Margin (M) as per Reg. Guide 1.99, rev. 2; the standard deviation for the (b) initial RT margin term is assumed to be zero since the initial NDT values were obtained from conservative (i.e., " upper bound") test RTNDT results. (c) Numbers in ( ). corresponds to surveillance capsule data. is cut into half, when Surveillance Capsule Data is used. (d) o3 1-3 0652D:10/063090

j 1 2.0 OESCRIPTION OF KEY ISSUES The reactor vessel issues that were addressed are pressurized thermal shock (PTS), low upper shelf fracture toughness, and operating limitations (i.e. heatup and cooldown pressure-temperature limits). All of these issues are associated with irradiation embrittlement of the reactor vessel materials. Figure 2-1 depicts the effects of irradiation on the charpy V-notch fracture toughness of reactor vessel steels. One effect is the shift in transition The other temperature from ductile (high toughness) to non-ductile behavior. effect is the drop in fracture toughness at higher temperatures (i.e., the upper shelf). A significant increase in transition temperature raises a PTS concern and causes a potential restriction in pressure-temperature limitations during heatup and cooldown of the vessel and plant. A low upper shelf impact energy raises concerns relative to both normal and abnormal conditions, including PTS events. Large transition temperature shifts are generally associated with weldments containing high copper content while low shelf behavior is also generally associated with high copper weldments, but only for specific types of welds. The PTS Rule [1] outlines regulations to address the potential for PTS events on pressurized water reactor vessels. PTS events are transients that result in a rapid and severe cooldown in the primary system coincident with a high or The PTS concern arises if one of these transients increasing system pressure. acts on the beltline region of a reactor vessel where a large transition temperature shift exists because of neutron irradiation. Such an event may produce the propagation of flaws postulated to exist near the inner surface of the reactor vessel, thereby potentially affecting the integrity of the vessel. Screening criteria for shifts in transition temperature for reactor ve;sel materials have been prescribed in the PTS rule such that operation above those criteria requires extensive analyses to defend safe continued operation. l 1 06520:10/063090 2-1

. i 10CFR50 Appendix G [3] contains a minimum upper shelf impact-energy requirement of 50 ft-lb. Non-compliance with this criterion requires a volumetric examination of the beltline region, additional evidence of material toughness from supplemental fracture toughness tests, and a safety analysis demonstrating continued safe operation. The Beaver Valley Unit I reactor vessel surveillance capsule exhibits a more than adequate shelf level for continued safe plant operation (HCAP-12005) [5]. Figure 2-2 gives an example of heatup and cooldown operational limit curves that were generated in accordance with the ASME Code (4), which is made mandatory by 10CFR50 Appendix G [3]. This figure also exemplifies how the pressure-temperature limits become more restrictive with increasing neutron embrittlement of the reactor vessel; i.e., the operating window between competing limits can become very small at the lower temperatures. Another underlying concern regarding the shif t in transition temperature is the uncertainty in defining the material chemistry that is used in predicting future transition temperature shifts and in the irradiation damage prediction formulations themselves. The following sections discuss the uncertainties that were taken into account in the definition of neutron flux reduction goals. 2-2 06520:10/063090

\\ Energy (f t lb) 70 80 Unirradiated irradiated so 40 30 20 10 II Il 11 o I' I -1 10 200 Tem ('F) 0 200 400 -200 Effect of Irradiation on the Charpy V-Notch Toughness (Sample) Figure 2-1. 0652D:10/063090 2-3

.s Pn Hootup or Costdown Curve at I.Nethne, W e Hoelee or,wwen., r Curve at t + At s t.oer sture Over-Preeeure (LTOP) Lhudts' 8afe 5tCS Pnsecure 't :_ _r - t.sved ResN to .....Needed.Acrees RCP No.1 Seal h onsor to see Pune I

I

• Adellenal AT remuning h deleys to plant'startup and *h Figure 2-2.

Pressure-Temperature Heatup and Cooldown Operational Limit Curves (sample)

3.0 IRRADI ATION DfBRITTLEMENT PREDICTIONS Neutron irradiation has been shown to produce embrittlement which reduces the toughness properties of resctor vessel steels. The decrease in the toughness properties can be assessed by determining the increase to higher temperatures of the reactor vessel material reference nil-ductility ' transition temperature (RTNDT). Because the chemistry (especially copper and nickel content) of reactor vessel steel has been identified as a major contributor to radiation embrittlement, methods hcve been developed to relate the magnitude of the increase in RI to the amount of neutron fluence. Based on the initial NDT RT value and the material chemistry of the reactor vessel limiting core NDT values are determined. region materials, the post irradiation RTNDT Westinghouse, other NSSS vendors, the U.S. Nuclear Regulatory Commission and others have developed trend curves and methods for predicting adjustment of RT as a function of neutron fluence and copper, and nickel content. The NDT two prediction methoris of most importance to the Beaver Valley Unit I reactor vessel are the methods used in 1) the Pressurized Tharmal Shock Rule [1] and

2) Regulatory Guide 1.99 Revision 2 [23. Currently the method identified in the PTS rule is required to be used for the evaluation of reactor vessels against the prescribed PTS screening criteria. However, a more recent method has been developed and is identified in Regulatory Guide 1.99 Revision 2.

This method is expected at some time in the future to be required to be used in the PTS Rule. Therefore, both of these prediction methods have been considered in the neutron flux reduction evaluation. 3.1 Pressurized Thermal Shock Methodoloav In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of reference temperature for PTS (RTPTS) at a given time. l are actually The prescribed equations in the PTS rule for calculating RTPTS F r the purpose of comparison with one of sevm al ways to determine RTNDT. for the screening criteria, which are discussed later, the value of RTPTS 06520:10/063090 3-1

~ the reactor vessel must be calculated for each weld and plate, or forging in is the lower the beltline region as given below. For each material, RTPTS of the results given by Equation 1 and 2. Equation 1: PTS - I + H + [-10 + 470(Cu) + 350(Cu)(N1)] f.270 0 RT Equation 2: PTS - I + H + 283 f.194 0 RT where I - the initial ref& ence transition temperature.of the unirradiated material measured as defined in the ASHE Section III Code, NB-2331. If a measured be used: 0*F value is not available, the following generic mean values mu: for welds made with Linde 80 flux, and -56*F for welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes. H - the margin to be added to cover uncertainties in the values of initial In NDT, copper and nickel content, fluence, and calculation procedures. RT Equation 1, H-48'F if a measured value of I was used, and H-59'F if the generic mean value of I was used. In Equation 2, M-0*F if a measured value of I was used, and M-34*F if the generic mean value of I was used. Cu and Ni - the best estimate weight percent of copper and nickel in the

material, f - the maximum neutron fluence, in units of 10 n/cm2 (Energies greater I9

'than or equal to 1 HeV), at the clad-base-metal interface on the inside surface of the vessel at the location where the material in question receives the highest neutron fluence for the period of service in question. 3-2 06520:10/063090

Note that the chemistry values given in equation 1 and 2 are best estimate mean values. The margin, H, produces upper bound RTPTS predictions.

Thus, the mean material chemistry values are to be used when available so as not to compound conservatism.

3.2 Reaulatory Guide 1.99 Revision 2 The Nuclear Regulatory Commission (NRC) has developed a method for predicting radiation embrittlement of reactor vessel material which is published in Regulatory Guide 1.99. Regulatory Guide 1.99 was originally published in July 1975 with a Revision 1 being issued in April 1977 and a current Revisian 2 (2) issued in May 1988. The Adjusted Reference Temperature (ART), bascd on the methods of Regulatory Guide 1.99 Revision 2, can be compactly described by the sequence of equations listed below: ART - Initial RTNDT + ARTNDT + Margin NOT - [CFlf(.28 J 0.10 Log f) where ART I9 f - Neutron fluence, n/cm2 (E 1 HeV), divided by 10 CF - Chemistry factor from tables for welds and for base metal (plates and forgings) (if no data use 0.35% Cu and 1.0% N1) The neutron fluence at any depth in the vessel wall is determined as follows: f-fsurf. (g-0.24X) x - depth into vessel wall from inner (wetted) surface Margin - 2 [o +o35 0 where og - standard deviation for the initial RTNDT. If the initial RT is measured, og is considered to be O'F. If initial NDT RT is not measured, o is taken to be 17*F for welds. NDT g 06520:10/063090 3-3

3 = stand &rd deviation of ARTNOT; 28'T for welds and 17'F for o base metal ex_ cept that o need not' exceed 0.50 times the mean 3 value of ART NDT* [. f 3-4 0652D:10/063090-

=_ 'I 4,0 PLANT SPECIFIC EVALUATION In the assessment of reactor vessel material conditions in accordance with the PTS rule [1] and Regulatory Guide 1.99, Revision 2 (2), an area of uncertainty is establishing the best c:timate chemical content of critical reactor vessel welds, particularly the copper content. Best estimate values of copper and values nickel content are needed for use in the equations to project RTNDT for comparison with the screening criteria for PTS and also to develop heatup and cooldown curves. Figure 4-1 identifies and indicatts the location of all beltline region materials for the Bea er Valley Unit I reactor vessel. The materials of importance for the beltline region are the intermediate and lower shell plates I and the associated longitudinal and circumferential welds. Engi,rized Thermal Shock Evaluatton 4.1 Pressurized thermal shock evaluations were performed using (1) PTS rule and (2) Regulatory Guide 1.99, Revision 2. These calculations were carried out for the entire beltline region materials for the Beaver Valley Unit i Reactor Vessel. These results are reported in Table 4-1. Also, plots were developed, j RT or RT Vs. fluence and are shorn in figures 4-2 and 4-3. These PTS NDT figures are applicable as long as i) the material chemistry of the beltline region remain unchanged, 11) the flux remain unchanged. i 4 l l I 4-1 06520:10/063090

-.... ~ -... -.. -. - 1 CIRCUMFERENTI AL SEAMS VERTICAL SEAMS 270' $6607 2 [ 19-7148- / f 8.2" 45 CORE l o. j, 180' { CORE E s 144.0" g 86607-1 19 71 ~ 90' y ~ C 9< f g 20.5* 4 11-714 = o 270* 20-71 1 87203-2 m 15 = CDRE y. 0+ 100* I 49.l* i l. 20-714A 36903-1 rigure 4-1. Identification and Location of Beltline Region Material for the Beaver Valley Unit 1 Reactor Vessel 4-2 06520:10/063090

• I TABLE 4-1 PROJECTED RT FOR THE BEAVER VALLEY UNIT 1 PTS REACTOR VESSEL BELTLINE MATERIALS RT PTS RT (Reg. Guide PTS Projected (PTS Rule) 1.99, Rev. 2) PTS Screening Component EFPY Fluence (1,2)

(*F) (*F) Criteria Int. Pit, B6607-1 32 4.07 217 214 270 -Int. Pit, B6607-2 32 4.07 247 244 270 Lower Pit, B6903-1 32 4.07 253 254 270 Lower Pit, B7203-2 32 4.07 190 188 270 Long. Held, 305424 32 0.75 173 186 270 Long. Hold, 305414 32 0.75 209 203 270 Circ. Held, 90136 32-4.07 198 190 300 Int. Pit, 86607-1 48 6.11 231 222 270 Int. Pit, 86607-2 48 6.11 261 252 270 Lower Pit, B6903-1 48 6.11 274 265 270 Lower Pit, 87203-2 48 6.11 204 196 270 Long. Held, 305424 48 1.13 192 208 270 Long. Held, 305414 48 1.13 233 227 '270 Circ. Held, 90136 48 6.11 220 201 300 I9 2 (1) Fluence is in 10 n/cm, (2) Projected fluence from Reference [5). 0652D:1D/063090 4-3 i

} A e v n gm S c 2 2 2 2 2 3 I ,1 h 3 5 7 9 1 3 5 ) 1 9 1 00 0 0 0 0 o 0 0 0 0 0 0 / / 1 F i ~ / p gure -/' / 4 / 2 l / s F lu F e E . / f', n L c U e V NC j s s / R E T 1 P T k /, S 0 6 M 1 L 0o f 0 0n E r s9 4 B l 4W e 9 3el l t i d / N in 6 / 6 j / e 0 C 7 Re M t g X i / J 2l m o X n y 2 /x / // / M a ) te h // j x r i 7 / [ a l s f ( P T S / R u 8 8 8 7 6 6 l 2 6 9 e o o o 3 ) 3 7 2 2 i

f.. ~

2 L u g yegg 310 / / (305414) j/ y 290 Long.' Weld-f / '86903-1 y (30s4243 7 a / / 270 / / / 866o7-2 y / ./ / / 250 // y /// n LL g39 / 7 '"irc. Weld. C y / p 210 ,/ / O g/V / /3/-^ 81203-2 9 9 d j) v t 190 p( f g / / n I h 170 / ) k / / 150 7-g // / / 130 / / ,s // / l10 / 90 0.I l.0 10.0 FLUENCE.'C10EI9 N/CMXX2) Figure 4-3. Fluence Vs. RT for Beltline Region Materials (Reg. Guide 1.99. Rev. 2) PTS

l 1 g j. 4.2 deltup_3nd Cool @wnlurr_yti Heatup and cooldown curves are developed for the Beaver Valley Unit I vessel based on the most limiting beltline region material. RT calculations at HDT 1/41 (T being the thickness of vessel at beltline region) and 3/4T locations were done for all the beltline regions. It was found that the lower plate B6903-1 is the most limiting material for the development of heatup and cooldown curves. RT values at 1/4T and 3/4T locations for 16, 24, 32 and NOT 48EFPYS are provided in Table 4-2. Figures 4-4 and 4-5 show the present heatup/cooldown curves which are applicable for 9.5 EFPY. Westinghouse generated heatup/cooldown curves for The inner surface peak 16, 24, 32 and 48 EFPY using Reg. Guide 1.99, Rev. 2. I9 2 fluences projected for 16, 24, 32 and 48 EFPYs are 2.112 x 10 n/cm, I9 2 I9 2 I9 2 3.168 x 10 n/cm, and 4.07 x 10 n/cm and 6.105 x 10 n/cm respectively [5). The heatup and cooldown curves for 16, 24, 32 and 48 EFPY are shown in figures 4-6 througn 4-13. i 4-6 06520:10/063090

o :. TABLE 4-2 RT VALUES AT 1/4T AND 3/4T LOCATIONS FOR 16.24, 32, AND 48 EFPY NDT FOR THE LIMITING PLATE USING REG. GUIDE 1.99, REV. 2 16 EFPY _24 EFPY 32 EFPY 48 EFPY 1/4T 3/4T 1/4T 3/4T 1/4T 3/4T 1/4T 3/4T ('F) ('F) (*F) ('F) (*F) (*F) (*F 215* 188* 229'* 199* 238** 206* 252** 218' (224) (179) (243) ('98) (264) (210) (270) (229) l I9 2 Projected neutron' fluence (E > 1 HeV) at inner surface are 2.112 x 10 n/cm, I9 2 I9 2 I9 2 3.168 x 10 n/cm, 4.07 x 10 n/cm and 6.11 x 10 n/cm for 16, 24, 32, and 48 EFPYs, respectively [5), values using chemistry factors based cn Number in ( ) represents RTNDT surveillance capsule data.

• For intermediate plate, 86607-2'
  • For lower plate, 86903-1 TA8LE 4-3 l

RT VALUES USED AT 1/4T AND 3/4T LOCATIONS FOR NDT DEVELOPMENT OF HEATUP AND C00LD0HN CURVES M-( _24 EFPY 32 EFPY 48 EFPY 1/4T 3/4T 1/4T 3/4T 1 '4T 3/4T 1/4T 3/4T ('F) (*F) (*F) (*F) ('F) (*F) ('F) I 224 188 243 199 254 210 270 229 i l 06520:10/070290 4-7

s [RIALPROPERTYBASIS op CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE B 6607-2 RT AFTER 9.5 EFPY: 1/4T, 202*f NDT 3/4T, 176'F CURVES APPLICABLE FOR C00LDOWN RATES UP TO 100*F/HR FOR TN DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUMENT ERRORS. UP TO 9.5 EFPY. 1 l l i ii 2500 ya o r ~ ,1-i _t lt l 7 l 1 2250 r i [ 2000 i I b I 1750 t ~-- / S 1500 L-' 1 T.' Unacceptable Operation / t I Acceptable l1250 y w j Operation 1 i g e / 9 1000 a N J. Cooldown, MDJ u 750 E Rates wr e oF/Hr s"'". r wem e e 500 0 d'" 20 1 4 6 250 i 0 50 100 150 ' 200 250 300 350 400 450 500 0 1 IMOICATED TEMPCRATURC (DCC.F) Figure 4-4. Beaver Valley Unit 1 Reactor Coolant System Coold (Current Curves) (No Margin f or instrument Error) l 4-8 06520:10/063090

k MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL PLATE B6607-2 RT AFTER 9.5 EFPY: 1/4T, 202'F NDT 3/4T 176*F t CURVE 5 APPLICABLE FOR HEATUP RATES U? TO 60'F/HR FOR'THE SERVICE PERIOD 9.5 EFPY. DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUv.ENT ERRORS. 2500, j 2i o iiiiiiiiiii, I J J Leak Test r i n c-Limit i [ f 2250 ? l l I I 2000 l r 1750 Heatup Rates 1 i ~~ UgF/Hr --f f To r 60 1500 l l a h i r r w g 1250 gaacceptable i I g Operation a r k1000 / Criticality / Limit Based a llf j on Inservice Hydrostatic / 750 f Test Temp. I / (329'F)for 4 f the Service Pt'riod Up Ti 500 2 9.5 EFPY Acceptable ~ 250 Operation t I f f I t t t i g 0 50 100 150 203 250 300 350 400 450 500 ledelCMED TEnsPCRatuRC (DCo.F) 1 Figure 4-5. Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations Applicable for The First 9.5 EFPY Using Reg. Guide 1.99. Rev. 2 (Current Curves) (No Mirgin for Instrument Error) 06520:1D/063090 4-9

4 s i I I ib O O ~YUI L I4 I i i l I ! (! t 1 4 I I i i i I I I i II , i i i! !!, I

iWi, i if "T~

i +"**- T-i

r i ~ t i

e. i l i j l i ' H i ! I i -. I 1 i.a i C 1,imiting 1/4 T RT . 224*F h lj ';ll* l l 7i ~~ 189'P 'T i,, ij ( j

3 Limiting 3/4 T RT 1

= M -t r-i~ ~- ~,- j i ?; iii i t~ r p i,

} ;I ! ! ' b-bbl H!

lI i i i i T;; i i ..,>i 2000! 4 i i ,j l l ', /l ll l l;, ,i,i i.i! ~ f~ i I* ~ i AI ii ii; H Ii i il 4 6 4 ii i,i r~i I '~ ^ i-+b 1750 / i 'i

i; j -

ii) ~~1 -7 i 1 i !. ii6 iiie i I i i / _. g i 'i i # i i,i i i i e ,. i. 'd_y - la__! 'i i i ! !t ii I r k ' dO ll l !/ i G 1500l ;';' ! i, i i

/

j 4 't Unacceptable i j ,, { e 6 T { Operation / I,4 i i i, i u 1250 il ,/ ,il}b

• ~

m.;; e / Y YI i i / ',q, 1 !(!

• I 6 *I

{ i, I I 6 i ! E l 1 e g i .s i 1000 M c. ! i i i ! i i ,i i /pr AcceptaDie 16 {T i O t i /# Operation ~Q /M/ 'i! i L, L, ! 4 ,, i .,. 0 i, ,yfjf; u 3 o Cooldown IMY/ I k. I M### # T-i i ' ~ 2 R.ates 1 1' / / -t-F/liR ,,g s - f i i 0 e 7,- ,f 500 ~~~ 20 .o 7 i -/ ~~ p L I 40 7 a _.,L i r i ! i i ~~ 60 i c50 goo ,j i i

• i 7i,-

i , ti!!ti L.1 ' ' I' 0 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPLRATURE (DEC.F) 1 Figure 4-6. Beaver Valley Unit 1 Reactor Vessel Cooldown Limitations Applicable for the first 16 EFPY U Ing Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-10 06520:10/063090 4

-~ ~- .t l" .i 1 L L.i L'_4 f !1 i ~ 2 { ~ u.h I l T sl [Y ~ i !!.I i i 4 ,,[ ,1 r,I - i, 1 H - Leak Test Limit f j

i, T~P Limiting 1/4 T RT

= 224'T T"- ~ ,L p.g 1((

- ;) ' I j

f 1 iT'TWM'T ~ * " ~ I I I 'i WT 188'F ~! ~~~T li ! _2 Iw

,,,,,,,, Pl F~*

Limiting 3/4 T RT a WI i t G 4 l'/:s, a e i,, -m. i ; ei mm .~ l W: I ' i i ! -! ! U alri

1 i */

. l r_ ~ ' --d ./ / i g;

• iT! !

Q + i ! i 'iiu! >t s! 1750[-t1.f'

d-l-k !

l# jf d-- [ , l[ l,IJ,' t 1 1/ [ i I i ~i~ ~ i T'Y' )1 '-1'T77 i it i i I 1i i g ~ G 153 0 F i, i i i / -- .T.i, ,m ptj,,iii ~

t-~t ;--.'

• t*?

unacceetabte / l' 2-H, '- r+H +-- c i i i ii / ./ i i i! i 'c;, T T T,'T iii i i i i i : F~~~~,Opcration i T' / / i i i y i w ,i i i i Tf i i ( / / i ! ii i t i if1 i< / f 7~ iii 'T-"p.ii_ y 1250 i ii i i l_M.~l t/[p-.) - 3 -Q ! r-ii6 -f, _~_ h.,i ~ i 4. ~_. /. Acceptable '~*~ "i iii ii .a _i _ _eq .!,1 't-Operation H-H* i i 1000 m _.p.11 C- ! ' ! '{ y' ' ! ' 1 l, p' p! ' i_60'F/HR L I-

{

t o g p- - y' l Fq _L y 9-j 7-7, i i r-i r e i M. 750 1-i - 1 j j j 7-- ,- t-7- a [- i iL/ 'I l' i tIjq ..~, II7 J . b !,j.3 k,,. V i -- h't LA-- 500 --I-- 2 I I-t t i I i I i iI ii! Criticality Limit Based on },.i.T' h L 1 l! l, 250 Inservice Hydrostatic Tasc i ,~ j;i Temp. (351*F) for the Service

• Period Up To 16 ErPY

-'~'~~'"1 i ii i T ii t-b-'f f-iII'!' I } ii i 0 O 50 100 150 200 250 300 350 400 450 -500 _ l,- INDICATED TEMPERATURE (OCG.F) 1 Figure 4-7. Beaver Valley Unit 1 Reactor Vessel Heutup Limitationc Applicable for the first 16 EFPY Using Rog. Guide 1.99, Rev. 2 (No margin for instrument error) 4--I A 06520:10/b63090 _ _ ~. --w

it s i 6 ii,i;i;ii(l i l" >,1 i i ! iii.) 'T i 50 3 j O'*U =,1 ii t 1

_- -..,i 1._,,: - -

~ i *-{ i T't-i-i r -* i. . r.i i ! i ; i i i ,,i6 -p ' tj'* h-k'hb'I

• ~*Y

'-~~~ , i -t.-dl f l 4 n.T w----- P-- rr-1.. I-1! ! ' ! !5M 1 r I I -l j, '}- )f l, l; l-l ' ', j i - { . m_ ' " ' k- - Limiting t/4 T RT 243'F L-. Limiting 3/4 T nrWI. 199'r il --+--m NDT L._ H-- / i 2000! i,, .,it i i P -ti,i,; i ji ) iii i i,ii '""-~ i / "T 7 ^i, fTT'I~ii!*I '-*-} 4 i ! i F i ,.;,4 iT i, i i !U2 ii (~l~I i i i i Ti I i i _ l_ / I'0 ff,! ! $ !h I I~I'? 1. ?Yt*? -b*i I IE~ ! i i i

• • • ~

'ti-p "'T*PT i i i, l

i i

i i i I"T;'--"' i 4 i ' + i /, i iiit4i,i J'" i i T t-P["-P ..i i .i i iiii 7"~ (-*, i i

i..i )

i i ( 3 i i i . +. - - T '.* T ' 2 '.500 ll -iYi l i -k) !.b i rrrrrr-n i; i, ~ * * " * - * " * ~~ ~" i Unacceptable ij j TT~a F-i* , i i ( ,ii 1250 ;; { , d."{.4 i T , Operation ,_]I ,,, [' ,i i +-{ y -,

; ;.7-q w-

- -pjj u j.

m..,

g p,.. _if ------. . ~. t.-t v7.4 3 _1..,.,-. o =.p,., f, } p....) -- ~+--Y ~ w w - -{...._-~ y g,H - H .L.. +- 7 iL z.7 u 2 1000 y-- (( j j A ~,. q f 5-Is i 1. a.;. p.{.{ J_LAL.-..l1 i.,J >i t 4 if 't t I gL o 1-. -l',I y yd }..

t..

- i+O., .-. x .a 4 -..+ p l T* 3cc,pg,g g, ~~'I-I' v s ,, d'_.b._. _ t i .u 750 ~ o eation o. --- Cooldovn j t l,,,L, p., ,3 i jj...q,,,,,, g,y,,,,2, .L *F/Hf 'aLQ .pse d,, 4,*f ( rat %8 e ,, j,, .q' ' ,-.11'.4 M~C@#f t '- AL c00 l-0 -it.*.'&,EIM-jh_.. -l,7 -.{.._.-.' 'L..,.p.. - jiTY. ~ 4 y 20 r,.p_.-.. ! !i q.. Ao na ng mm r r4-r++ y.t- .. rr so ~ rrwr r.1-. 4 r-y.g, 7 - - 100 H;; c.a o t.+p 7- + - t..- ~Y -}-p "I 7 ~ 3 _.... T* i [.9 i ~., - 7 1 7.,. ....p _1 4'.' v 1..i i t. T.. ' e LL L-L -- ~L l_ L-L -I. I -.0 0 4SO 500 0 L' t 0 50

• 100 150 200 250 300

%Q INDICATCO frumERATURC (DEC.F) i FQura 4-8. Be?.ver Valley Unit 1 Reactor Coolant System Cooldown Limitations Applicable for The first 24 EFPY using Reg, I Guide 1.99, Rev. 2. (No margin for instrument error) 06520:10/603090 4-12 I E w.A .M

~ P %9 ! LL.i T, T---*--] i!i-r T-~ TT_ ' i_i.l..L,.IQT ' T TT7 T i 6 N O - CIT.H m i ] i" I i.,i i _ I ~ - - i;1I. O! T 1.L.i ' ' L L4 i.i. J.[.d.1. _ 4J 'I'~~ Lar.,k Tenst Ltait -_,4, _4 i i, N.5 0 -i-N ', l Mk-b h---[ -"-* -*i-* t l HT i-7- +- ** -f ,i h-, ~-' "- 7 Lisiting 1/4 T RT

• 243'F

~ -t-' "M'f T*E] . - ~* 2."1 b Limiting 3/f. T U = - 199 ' F yt7

-- q h

JDT 7-~ ~

3. !** '

T tT, y -{y T/i. 3 -. i r4 "I .] d3]9 ii ,ri, vi,.

-ittt--H !.

t- . 3 i ~ TT 7j i--'-- i i;i i i.i i i-, *.., i i i i,4-+# ji,i i i _. m T p+ i iy .p.a. ,,3 i~,t.[I Q- $iI { I! kt-bM-b-N Y N' Tid i i -E i iN ,, 1--" - -

4.,

l0 iii 4 r*T7---",- ;, i, ["*' ,t ,i,e ~ i 1T*T-*' T.. i <3 T-' -f.1 i iL.r p. T 4 T. T T.. 1 i ,. T"-* r,, i,- i i

-t ~-, - T-

. t T t- ,1-i ;- tT-" l,lii )~r~~hi-b-'-i-f-k! Yk f

i 7

l ,~/*, ! 5 ( -t 'hb i-~" i Tt-r - r - vi i q-h T~_.T. h..'.i.~_,. ?t 4, i v y. - _t_1 .nr

sco

.-,i o fi - -t-* t.i, _, r c i itf I l l l. 4 Unacceptable -h-h.-- O ! ', i 9'd .m o I ~YM-'~ T T r -i ~ r s s' Ugetation 4 Y[ h...t t--F.. t--] L < 9, .. =.- Q.. =.=..7-u I P.50 v.L ~ 1., W. ,,,,~:,,_g,7, ct ,, g i

• p' P.Ti --

,,,,, g , ;/. H 1 ~ ' M i h t + b 1 R 4 ,g j..i/Trh} "a-- , hi -rr- ) _. , i,h- ' ir " M i i g-*- t1 c, 1000 A[..:-- Acceptable h -i '- f h-fp-':21.p. O 1 iF -: o eraua v .g a

. 6o'r/n i,tt.y.h7;i,ll!

p .-,@.p;M r O 750 [. ' by - q .{. 7, i i ri !, ! g Try r z ci .I I l T!_4 }:h-!.1 !l i Jc"L. Q' r._ I I -- - 1 C 500 O -t_i, ;p ~ t I 7, ,.1 !. 1 i M-i3 l a t 4'ii' . ) ~7",'T' Critica11r.y ).init Based en t ii i ~ t"** ##i"* UI # 't *ti" Tt i i d "~~ ' f

. Teirp. (369'P) for the Service "SO

.) i .L Period Up To,24 ETPY i i', F 1' i 1i ii l 'i i' i 'III'''''''''I i --IilI 0 50

• 100 150 DOG 250 300 350 400 450 500 0

INDICATED TEMPERATURE,(DEG.F) 1 Figure 4-9. Beaver Valley Unit 1 Reactor Ccalant System Heatup Limitations Applicable For The First 24 EFFY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 0652D:10/063090 4-13

7,, '. a a sa. 1 I Ft: ii

m.,

i i, ee i ii,ii c : 4,,

. c. a. 3.;i,.,,,

.i,iT: i i iiiii '-~ ~~~"* My' ? - a, i ;- h. ) i r i , 4, r' . F j i ** i i, ~ ' h-a - ri 'Q,J i F? i t i i i i i.., mr n t if i i,

P

-[ Limiting 1/4 T RT = 254'F i ) i ie!!.g Q .;.? LirJ ting 3/4 T RT' = 210*F / i r i 93 u. >3 / r i.t i_

• - 4

_. g~# 1I ! I 1

• I a

,4 c, i t. g***UT*

• g 4 i a i

- - I i 4 i f t 4 e v i 1 i i i ! - r7 t rm , 50 p :rr, ,-r, ri i e i, i-,74 3 ---r-i ,i r c, i 0 ,,ii i p.-t ri-m r h, i 4 I i d i . r"ITT 6 ii

• i F.

Q! 1 i + L 7 r.a.r. 4 i r r 1 i 7-kg ',[ l m { 1500 -Wftp i-fr- --M e ' ' F J.. ! .h,'-. Unacce ptable l ', ' N-ht h Operation .--g---- p 6 t-g y 1250 1 s i i i

• • 1' r,

o d-i i / i 6 6 i

• 6 l I 4 4

i J l i . i e s a i t-

ec...10 0 0 -,j

~1 .y i [ o A F J i i, 6

• u

/25 I ( I t ]. -5 h= '! d -- u ~~50 M-- h ) Acceptable M-- $ Cooldown j*g t,,%yf, 7 Operation.

'-h 2

i t1 Rates 1i 1 n # 69' / e i 'i4 I I i T' ~ 1 'F/R 1 500 ,A 0 u. _##1 7 o j 20 j m I* ~""" ' b 40- ~ 50 -i. 60 2 -7 100 y d' 0 ' ' ! U I. O 50 100 150 200 250 300 350 400 450-503 INDICATED TEWPERATURE (DEC.F) 1 Figure 4-10. Beaver Valley Unit 1 Reactor Coolant System Cooldewn Limitations Applicable For The First 32 EFPY Using Reg. -Guide 1.99, Rev. 2. (No margin for instrument error) ~ E '06520:10/063090-4 14 y

t

--e ~ y -a+m

~ ~ I 1 t bbb0 i L.h2:L 9 1 i i i i ii i I ! ? i ! I i ! ii i !l! I1 i li i I ! 6 i 6 ii! P i i ! i ii i i i i i i i i! I L4 4 ![! ! [, ' ' [ ; ' I Leak Test Limit L4 7 .'254'F l flj !,[ '[''[l' Limiting 1/4 T RT Liniting 3/4 T rig = 210'T i ,,f g i

f-j^

NDT , f. ,I IJ il It i ii $ i i , i, ; f ,i , ;i ,i j d300: i i + 1 i_ i! I i i t i i ! 4>><t i I i, ,iiiii j 4. j i i I . /l I /I I

• i ii i

t t/ /l i + i i, i! i ! ! i j / l!le, l 1"'50 [' {!j !'i(( / N ./ ! I ! ! ? i t 6 t iie i I i 6 iiii ._ I i a. i ! l / fi ii i { l t t i i it u / / T~I ! 6 i i 6 6 1 t i i i ! ! i li i E I i 31 1500 i t i, / '/ t i, ] ] ~ 'T~T~i". iii j ,j!, 3 !I J i o r l i i i ll Unacceptable j j Operation 7 i i_ i i I '/ i d 1260 i i i a ) i i ie i 1 6 I i IJ * - 1 i '/ i I ' 4 "O i 9 j k i r I l J e i 1 l 2 1000 ', {'i / l'! y Acceptable r O !<i) F't 3 i i i / Operation i, i i ii, i 6 i i+ / i ia i 6 7 / { i u 750 -Q 60'F/IIR / 9 / i t I i IM ii t i, ,i i 500 g ,,,,,-r i i l' l l Criticality Limit Based on ( Inservice Hydrottatic Test 6 )j j i Temp. (380*F', f or the Service 4 250 Feriod Lp to 32 IFFT I i Ii i I i l* I i ' 0' 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEWFERATURE (DEC.r) 1 Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations Figure 11. Applicable For The First 32 EFPY Using Reg. Guide 1.99, Rev. 2. (No margin for instrument error) 4-15 j 06520:10/063090

e I L ! i1 !! 1 iv- _.u:Lta ii i i i 1 I Ii !I I i i -+ ,p i.

Ji, t,

+-- 1 i !! ! li /iiii i 1 l i ' ' ' C F* I ,j~ H Limiting 1/4 T RTgg7 2707

i 1,

j i a Limiting 3/4 T RTNDT " 229 F I I !

1. I

~ i I. i 6 li 11 ! a r1 ~7 t II 1. ! . t 2 T 4 i ! ! l ;q i i i _l[ 20 00 f* - .i i j i i ji !e 17 t i i _i i i i .! I. l i iii i,i i i.i 11 ! i i ! t i j~F i!- i I!It !Ii I I ',^, ' ~5 3 - l-- Iiiit l;i i(-ti j i !T'i. I i iI i i i1 i i ! i ii / iii i Ii' i ~ ,jii i i I i ii j I ii!! i1 i I i 1i' i i; 3 1500 ri-+ ,i i-t _ i_ ii j l Ii _L 1 i ii! r ~ i,ii i / 1 i. 6,i r j i l [ w j ~ U"*C:'Ptable f i u 1250 Operation ti ji; !/ I t i u j ( E 1000 c,',; E

• i

-l-7,j[ i o d i 1W/ i i ! i i ' 4 i sMO i 3 750 cFr L_, -/,j[ Acceptable 9 !I I i 7---- 0 - Cooldown 1 E*F Li r'.TI g/ j coc 2 Rates i-i n-gg

y

• rntn

~ ~ -

O m- , [ y > #r ~ 20 1 250 40 i.. 60 ~~"" ~_ 100 !Ii[i i 5 i I 0 50 100 150 200 250 300 350 400 450 500 0 INDICATED TEWPCRATURC (DEC.F) 1 Figure 4-12. Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations Applicable for The First 48 EFPY Using Reg. Guide 1.99. Rev. 2. (No margin for instrument error) 4-16 06520:10/063090

t l l I 2500 ~5 I {,8 )jl fp i} 'IIIII IIIII IL4i!4 ij 3 ij i, ,i i _[ lli ! i I I 'I I Leak Test. Limit j l.1 A bl ;i -O { Limiting 1/4 T RT,GT = 270'T = i !I ii 1,_i_1 y Limiting 3/4TRThDT 229'F / l fiii'a 3 2000 - i I !Ai !i i h ( i ^ l ll i / i 4 i i i i,, I'-~ -/ I 1750 4 i / / l' ll.__., --+ i ii f / i ii j j i il G 1500 l / if !i i, J J !i / / 1 { f ~~~' Unacceptable --- j j u 1250 _ operation f i I } I / A l r Acceptable U E 1000 ~ / operation i n 0 f s / 750 so*F/HR a / Z f I I n 500 e j -Criticality Limit Based on Inservice Ilydrostatic Test Temp, (396*F)48 ETPY for the service 250 Period Up To ~ -j iiiiiiii!1i iiiiiiiiiii 0 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEWPERATURC (DEC.F) 1 1 Figure 4-13. Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations Applicable for The first 48 EFPY Using Reg. Guide 1.99, l l Rev. 2. (No margin for instrument error) 06S20:10/063090 4-17 l

.l 5.0 FLUX REDUCTION GOAL Neutron flux reduction goals for the Beaver Valley Unit I reactor vessel were established for the licensed life and an additional 20 calendar year life considered for life extension period for the key issues related to neutron embrittlement. The issues considered in this report a"e 1) Pressurized thermal shock,11) Operating limits and lii) Emergenc/ response guidelines. Flux reduction goals were established based on these three issues. 5.1 Elytnte_LimLt1 In order to set neutron flux reduction goals, a target end-of-life neutron fluence must first be established. In setting the target end-of-life fluence, a variety of key issues were taken into consideration to ensure setting the correct-target. Heutron fluence lin.its were established for both end-of-design life an'd life extension. The end-of-design life was assumed to be 32 effective full power years (EFPY). The life extension was assumed to be 48 EfPY, 5.1.1 Etessurized Thermt]_ Shock Criteria The consideration in setting the neutron flux reduction goals was pressurized thermal shock. PTS was assessed using the current PTS rule and the Regulatory Guide 1.99 Revision 2 methodology. Emergency Response Guideline limits (6) that have been established for operator guidance during PTS events were also evaluated. ?TS Rule Methodol.03y The neutron fluence at which the FTS screening criteria are reached was calcu-lated by solving both Equations 1 and 2, as described in Section 3.1, for f. I 0652D:10/063090 5-1 1 l

that is, RTPTS ~ I ~ N 1/0.27 l Il"I 10 + 470.Cu + 350 Cu Ni 4 4 RTPTS ~ I ~ N 1/0.194 3 f2 "'E 283 where RT is defined to_ be the appropriate screening criterion. The PTS screen-ing criterion as specified by the PTS rule, is 270'F for longitudinal welds and base material.,for circumferential welds, the screening criterion L is 300'F. i By using the material' properties for the beltline region and the appropriate margin terms, two neutron fluences were calculated. The lowest value of the two fluences is the applicable fluence. From Table 4-1, it can be seen that: the limiting material for the PTS; is the lower plate B6903-1. All beltline region materials are within the PTS Screening criteria (See. Table 4-1) except the. lower plate B6903-1. Flux reduction is required only for the plate values within the screening criteria. The above B6903-1 to maintain RTPTS methodology is used to develop flux reduction factor required for 86903-1 and is shown.in figure 5-1. Reaulatory Guide 1.99 Revision 2 Metho #1olv i RegulatoryGuide1.'99R$ vision 2 presets 1he'latestmethodthathasbeen ~i -developed for predicting radiation embrittlement which may in the future be required to be used in-the PTS rule. To account for this possibility, target. neutron fluences at.which the current PTS rule screening criteria will be reached were established in a similar fashion to the PTS' Rule Methodology of r the previous section. L L 06520:10/063090-5-2 ,*vs, ---twrv--y ,,v..<wyww-,-w- ---w ,wp m - = y-w wrw-v ywre-,-,ey--ww-+rw,vc.ipm,-- e+-.,3-+mysw -*+ver-i -<W y4 i s w. . wri wr -e---~ -t w--wmiT w w-e

For this particular case, the limiting PTS material is the lower plate, B6903-1. For the purpose of PTS evaluation, survell!ance capsule data are not term is used. And, since surveillance capsule data is not used, the o3 not cut in half. ART - Initial RTNOT + Margin + CF x Fluence Factor Rearranging those equations, ART - initial RTNDT - Hargin fluence factor - f(.28 .10 log f) CF By using the appropriate chemistry, margins, initial RTNDT, and the PTS screening criteria values, a fluence factor can be determined. The neutron fluence may then be determined from a plot of the fluence expression, which is presented in the Regulatory Guide 1.99 Rev. 2. From Table 4-1, it can be values for all the beltline region materials are within PTS seen, that RTPTS Screening Criteria using Regulatory Guide 1.99, Revision 2. Hence, there is no need to develop any flux reduction factor curve using Regulatory Guide 1.99, Revision 2, 5.1.2 Heatup and Cooldown Curves Heatup and cooldown curves were generated for Beaver Valley Unit 1 Reactor Vessel for 16, 24, 32 and 48 EFPYS. These operating limits are shown in figures 4-5 through 4-13. Using these data low Temperature Over Pressurization (LTOP) system set points were developed (7). Detail discussions xere held with Duquesne Light Company personnel to investigate the consequences of these operating limits on the operation of the plant. The LTOP system set points were thoroughly scrutinized >3d it was agreed upon that the operators will have sufficient margin for the safe startup and shutdown of the plant. Na flux reduction is required for the operating limits. 06520:10/063090 5-3

f ^ 4 9 x N Rn d CL a LL D i W y Rv e t 4 a j ~ u n b o O l &T~ b 8z 5 W oN 2 a x D E 1 o oO g Q ~ t la b a C D-j- ~ \\ 8 108d NOIlanOBB Xnld 038Inegg 54 _. _. - _. ~,. _ _.. _ _ _. _. - _ - _ _ _

Figure 5-2 indicates that at 180*F (48EFPY), the LTOP set pressure is 470 psig, based on a steady state heatup and cooldown allowable pressure of 548 psig (See figure 4-12). To maintain the Reactor Coolant Pump (RCP) Seal, the Reactor Coolant System (RCS) pressure must be maintained at above 325 psig before starting the RCP (for Beaver Valley Unit 1). Assuming an aoministrative margin of 50 psig, the LTOP system must be set at a Reactor Coolant System pressure of 375 peig for protecting the RCP seal. Another point of interest is, how long the vessel needs to be radiated, so that the LTOP set point drops to 375 psig at 180'F. It has been calculated that the steady state heatup and cooldown allowable pressure at 180'F will be 455 psig to have a corresponding LTOP set point of 375 psig. It has been observed that the Beaver Valley Unit I reactor vessel needs to be radiated for a long period (beyond 48 EFPY) to have a steady state heatup cooldown curves of allowable pressure of 455 psig at 180'F. This is because

1) the radiation damage becomes saturated beyond certain accumulated fluence (see Trend Curve of Ref. 2) and 11) the reference fracture toughness, KIR, see I'V'I' Ref. 4) dces not change significantly beyond certain RTNDT 5.1.3 Emergency Response Guideline Limits l

Emergency Response Guideline (ERG) pressure / temperature limits [6] were developed in order to establish guidance for operator action in the event of an emergency situation, such as a PTS event. Genaric categories of limits I were developed for the guidelines based on the limiting inside surface RT The following table presents the generic categories, which were NDT. conservatively generated for the Westinghouse Owners Group so that they would P be applicable to all Hestinghouse plants: ERG PRESSURE-TEMPERATURE LIMITS ERG Pressure-RTNDT Value Applicable Held Temperature Limit Applicable CATEGORY I RTNQT<200'F Longitudinal & Circumferential CATEGORY II 200 F < RTNOT < 250'F Longitudinal & Circumferential CATEGORY IIIb 250'F < RTNOT < 300*F Circumferential Only 1 06520:10/063090 5-5 l

for which the generic category ERG pressure / temperature The highest RTNDT limits were developed was 250'F for plate, forging and longitudinc1 weld and 300*F for a circumferential weld. Thus, if the limiting vessel material for a circumferential exceeds 250'F RT for a plate, or 300'F RTNDT NOT weld, plant-specific ERG pressure-temperature limits must be developad. For the Beaver Valley Unit i vessel, the limiting material is the lower plate, B6903-1, so the screening criteria for the ERG is 250'F. RT at the inner surface for all beltline region materials are shown in NDT Table 4-1. It indicates that Beaver Valley Unit i Vessel, plate B6903-1, will exceed the screening criteria of the ERG for plates at 32 EFPY. There, it is suggested that a new ERG for Beaver Valley Unit I should be developed before the vessel reaches 32 effective full power years. l l {- 0652D:10/070290 5-6 men-:1'- --+ - ~ - + w e---we ,-m-- p-r- -erwm,+ -, - w-m<---- ~, v,ww-- w-*me w-yrwm--pwww,-e g-m--

,t t t u .s. 7.. w \\ t - Ja C3 @g Q M ed IHliEl in _. 73 '; g 9 -^)) ~ x ("1 <w-..." ' ".. ' - = C. n a 1 _; .a i e Yk f"] IIl [] C '1 'O t i s I Figure 5-2. Acceptable LTOPS Set Point Range Vs. RCS Temperature 06520:10/063090 5-7

6.0 CONCLUSION

Ertt1Dille d_.llLtIMLiho t L.CLli tLh from Table 4-1, it can be see.. + hat no flux reduction is needed for vessel life of 32 and 48 EFPYs using Regulatory Guide 1.99, Revision 2.

However, flux reduction is needed for a life attainment of 48 EFPY using PTS rule.

The flux reduction factor curve is developed and is shown in figure 5-1 using PTS rule. Fuel management options may be considered to reduce the cumulative values for the fluence for the vessel life up to 48 EFPY, to maintain RTPTS Beaver Valley Unit 1, below the PTS screening criteria. However, it is to be noted that the NRC recently issued a proposed values. This methodology is methodology [8] to calculate the RTPTS essentially the same as the methods used in Regulatory Guide 1.99, Revision 2. This report identifies that the Beaver Valley Unit i vessel will not exceed screening criteria using Regulatory Guide 1.99, Revision 2. litAltup_And_.CDDldOMLCMIyei It has been concluded in section 5.1.2, that the Beaver Valley Unit I has acceptable margin of operation up to 48 EFPY provided 1) the Reactor Vessel Core maintains the current level or less flux 11) no significant change in material chemistry in the beltline region and 111) no significant change in the regulatory arena. E!!!E.IgencyResponie_Guidelinei The limiting material for Beaver Valley Unit i Vessel, has been found a be the lower plate B6903-1. The screening criteria for plate is 250'F for Emergency Response Guidelines. From Table 4-1, it can be seen that, the RT at inner surface up to 32 EFPY is 254*F and 265' up to 48 EFPY (using NDT Reg. Guide 1.99, Rev. 2). It is suggested, that a new Emergency Response Guidelines should be developed for Beaver Valley Unit 1 before the vessel reaches 32 EFPY, 065?D:lD/070290 6-1

7.0 REFERENCES

1. U. S. Nuclear Regulatory Commission, 10CFR50, " Analysis of Potential Pressurized Thermal Shock Events," Federal Register, Vol. 50, No. 141, , July 23, 1985. 2. Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," USNRC, May 1988. 3. U. S. Nuclear Regulatory Commission,10CFR50, Appendix G. " Fracture Toughness Requirements," Vol. 48, No.104, May 1983. 4. ASHE Boiler and Pressure Vessel Code, Section III, Appendix G, " Protection Against Non-Ductile failure," 1972. 5. Yanichko, S.E. et. al., " Analysis of Capsule Westinghouse from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Hestinghouse Electric Corp. Report HCAP-12005, November 1988. 6. Emergency Response Guidelines - Revision 1 Hestinghouse Owners Group, September 1, 1983. 7. Beaver Valley Unit I tow-Temperature Overpressure Prevention System (LTOPS) Setpoint Analysis at 16, 24, 32 and 48 EFPY, Letter Report DLH-90-528 January 23, 1990. 8. U. S. Nuclear Regulatory Commission, Proposed Rule, 10CFR50," Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Vol. 54, No. 246, December 26, 1989. l { 1 06520:10/063090 7-1

d ATTACHMENT F Beaver Valley Power Station, Unit No.1 Proposed Technical Specification Change No.191 Beaver Valley Unit 1 Low Temperature Overpressure Protection System (LTOPS) Setpoint Analysis at 16,24,32, and 48 EFPY 'I l t 9- -w -e www w -www-,-wer-. e- --+--------w. w-- w m-w r -+y---- w----w-wr-w-------

34(d-90 -J.18 d TABLE OF CONTEllTS Section Paceit References............................................................ 2 Introduction.......................................................... 3 Summary of Results.................................................... 4 Appendix G Limits.................................... ................ 6 LTOPS Set poi nt Generati on............................................. 6 (Appendix G limits)..................................... 11 (Overpressure-Setpoint Correl ation)..................... 17 l [ l l i 1 ~

Qlld JQ V' s References

1. Westinghouse Project Letter DLW-89-519. "LTOPS Setpoint Design Changes", 01/31/89

WU-90 .J~a p' 9 t Introduction The Low Temperature Overpressure Protection System (LTOPS) limits pressure i transients during cold shutdown, heatup, and cooldown operations in order to minimize the potential for imparing reactor vessel integrity when operating at or near vessel ductility limits. The imposition of this con-straint, together with the requirements for reactor coolant pump operation create a set of both high and low temaerature dependent 3ressure bounds. It is a regulatory requirement that tie upper bound not se exceeded. Viol-ating the lower bound is not a regulatory concern, but does result in damage to the reactor coolant pump number 1 seal. The goal of setpoint selection should be to prevent either bound from being exceeded below the 0 LTOPS enable temperature (275 F for Beaver Valley Unit 1). Westinghouse Electric Corporation was advised by Duquesne Light Company that the 9.5 Effective Full Power Year (EFPY) pressure vessel temperature limit (Appendix G curve) applicable to the Beaver Valley Unit I reactor vessel, and the minimum pressure required to start the reactor coolant pumps, has created an operaticnal constraint which is unduly limiting the rate at which the plant is able to heatup from cold shutdown. A meeting (May 11, 1988) was subsequently held at the Beaver Valley site with DLCo personnel in order to provide background and to explore possible solutions to the problem. As a result of that meeting, and a second meeting (July 14, 1988) held at the Westinghouse Energy Center, instruction was given to Westinghcuse by DLCc which defined the kind of analysis that would best meet the requirements of Beaver Valley Unit 1. The results of that analysis were provided to DLCo by reference 1. Subsequently, DLCo expressed an interest in obtaining LTOPS setpoints for several additional vessel exposures; extending to 48 EFPY, These addition-al setpoints were generated from steady state pressure temperature limits based on revision 2 of NRC Regulatory Guide 1.99, as applied to the cal-culation documented by reference 1. I i 3 i ~ r ~

M S Ju F ^ Summary of Results The result of the analysis is summarized by Figure 1, illustrating the range, as a function of RCS temperature, of acceptable LTOPS setpoints for steady state pressure temperature limits based on revision 2 of NRC Reg. Guide 1.99. Tne figure also shows the reduction in setpoint range as a function of increasing reactor vessel exposure. The result of the analysis is that, at low temperatures, the reactor cool-ant pump number 1 seal will not be protected. The calculation requires the assumption that one of the pressurizer power operated relief valves (PORV) i dedicated _to the LTOPS function will-fail, resulting in a relatively high-er overpressure.-Consequently,_the analysis indicates that no combination of setpoints exist that will preclude opening both PORV's. The result of both valves opening is a pressure undershoot much larger than that exper-ienced fnr just a single valve opening. It should'be noted, however, that the algorithm currently employed by-Westinghouse is quite conservative, in that the maximum possible charging flow (typically, in the 400 gpm range) is assumed at the LTOPS setpoint selected for the parameter study. In 3 reality,- the charging flow probably will not exceed 150 gpm, considerably reducing the resulting overpressure. At low temperatures, it is recommend-ed.that OLCo take advantage of this conservatism and select the setpoints-such that the differences between the setpoint pressures are maximized. At a v ssel exposure of 9.5 EFPY with a reactor coolant system temperature of 80 gF, for example, select the first opening valve setpoint at 400 psig, and the second at 440 psig. As the reactor _ coolant system temperature increases, the margin between the Appendix G limit and the reactor coolant pump no. I seal limit also increases. Eventually,-a temperature is reached that comfortably allows the selection of staggered-LTOPS setpoints-which will preclude opening - both PORV's;-even with the high-mass injection rates assumed by-the analysis. From Figure 2.13 of reference 1, at a reactor coolant system 0 - temperature of 120 F (vessel exposure of 9.5 EFPY), the first opening . valve can be set at 395 psig, and the second opening valve at 475 psig. As vessel exposure increases, the temperature which allows these staggered setpoints also ine.reases. At 48 EFPY, for example, the minimum temperature 0 Table-1 at which staggered setpoints can be obtained is about 180 F . sumarizes, as a function of vessel exposure, a set of recommendeJ set-points which' preclude multiple valve opening, and-the estimated minimum temperatures.at which these setpoints can be implemented. Table 1: LTOPS Setooint Summarv Reactor Vessel Reactor Coolant Setooint (osia) Exoosure (EFPY) Temnerature (O l Valve No. 1 Valve No. 2 f 9.5 120.0 395.0 475.0 16.0 136.0 390.0-470.0 24.0 152.0 390.0 470,0 32.0 162.0 390.0 - 470.0 48.0 180.0 390,0 470.0 4 S'4"" 4 4 Tevr

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Q4A Qo- & M Anoendix G Limits The steady-state cooldown Appendix G 1 aits, without instrumentation uncer-tainty, were developed by the Materials Technology group within Westinghouse (reference Attachmement 1), and form the basis for the LTOPS setpoint selec-tion. The limits were generated for reactor vessel exposures of 16, 24, 32, and 48 EFPY, LTOPS Setooints Generation The reactor coolant system pressure extrema, shown by figure 2, is repro-duced from Table 2.7 of reference 1. The reference 1 calculation considered mass injection events only. The setpoint evaluation for these additional exposures is similarly based. The correlation between the Appendix G limit and the maximum setpoint pressure was obtained by assuming a linear rela-tionship between the maximum overpressure and the setpoint pressure: Setpoint Pressure ao + a3(Max. Overpressure) By the methoo of least squares (reference Attachment 2): o 105.133 a ag = 1.04821 The maximum LTOPS setpoints, as a function of reactor coolant system temp-erature, were generated by substituting the appropriate Appendix G limit for the 'Hax. Overpressure" in the above equation. These setpoints are provided by Table 2: l Table 2 DLW Maximum LTOPS Setooints for the Indicated Accendix G Lim,il Vessel Exposure (EFPY) 16.0 24.0 32.0 48.0 Temp (O ) Limit Setpt Limit Setpt Limit Setpt Limit Setpt F 85.0 516.6 436.4 509.6 429.0 506.3 425.6 502.2 421.3 100.0 523.8 443.9 515.1 434.8 511.1 430.6 506.0 425.3 140.0 553.1 474.6 537.6 458.4 530.2 450.6 521.1 441.1 180.0 605.3 529.3 577.5 500.2 564.3 486.4 548.3 469.6 220.0 698.3 626.8 648.9 575.0 625.4 550.4 596.6 520.2 247.1 800,0 733.4 260.0 776.0 708.3 734.2 664.5 682.9 610.7 265.5 800.0 733.4 276.3 800.0 733.4 280.0 748.7 679.7 292.3 800.0 733.4

1. Indicated pressure in units of psig.

6

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QLD 9 0 - dc V' The setpoints from Table 2 are rtriauced, with roundoff, on Table 3. In addition to the maximum allowed LTOPS setpoint, the table provides the mini. mum allowed setpo'nt, for the RCP Ho. I seal protection, and the differcnce between the maximum and minimum setpoints. This difference orovides the range, as a fur.ction of reactor coolant temperature, from witch the.'10pS setpoint(s)c'enbeselected. The maximum LTOPS setpoints, from Table 2, are graphically shown as a function of reactor coolant system temperature, parametric with vessel exposure, on figure 3. The acceptable setpoint range is indicated to the right of the bounding curves. The 9.5 ErPY limit has been copied directly from figure 2.16 of reference 1. Note that the " notch" accounting for the reactor vessel flange ligaments has essentially disappeared for vessel exposures in excess of 16 EfPY. The minimum system pressure required to protect the reactor coolant pump number 1 seal is unaffected by neutron fluence, and thus remains unchanged from reference 1. r l l h l 8 I we wrer-w e v y--r rw-.---r:-.-,-- ,e-ev.,..- .re..--c-ew,,,- -,. --- --r-~-.r-

Table 3 DLW Maximum LTOPS Setpoint Spread at the Indicated Aprendix_G Limit Vessel Exposure (EFPY) 9.5 16.0 24.0 32.0 48.0 RCS Temp Min Max Max Max Max Max (Deg F) Setpt Setpt Delta Setpt Delta Setpt Delta Setpt Delta Setpt De i 85.0 396.0 444.0 48.0 436.0 40.0 429.0 33.0 426.0 30.0 421.0 25.0 100.0 394.0 454.0 60.0 444.0 50.0 435.0 41.0 431.0 37.0 425.0 31.0 120.0 392.0 472.0 80.0 457,0 65.0 445.0 53.0 439.0 47.0 432.0 40.0 140.0 389.0 496.0 107.0 475.0 86.0 458.0 69.0 451.0 62.0 441.0 52.0 160.0 388.0 529.0 141.0 498.0 110.0 476.0 88.0 466.0 78.0 453.0 65.0 180.0 386.0 572.0 186.0 529.0 143.0 500.0 114.0 486.0 100.0 470.0 84.0 200.0 385.0 629.0 244.0 571.0 186.0 532.0 147.0 514.0 129.0 491.0 106.0 G) h Notes: Min Setpt prevents RCP No. I seal closure (;, i Max Setpt prevents exceeding the Appendix G limit -C) o ^ Setpoint pressures in units of psig r kc R

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_....._._.._.._7.___...-._. ._ _...... _......._. ~ 'hy QO-jay acy e HT SMART 177-(89) MATERIALS TECHNOLOGY a 236 6465 Ori November 3, 1989 gr Data Points for Developir.g LTOP Set Points 9 for Beaver Valley Unit i Vessel i s..e= 9J. P. j,ut'z -f('//J/77&'g('.., to cc: N. P. Hueller D.-C. Adamonis .T. A.'Meyer 'S. S. Palusamy T. R. Mager R. D. Rishel As a part of Reactor Vessel Operating limit study for Beaver Valley Unit 1, Structural Materials and Reliability Technology has completed the generation of heatup and cooldown curves. The following information are attached for your use in the development of LTOP setpoints. l 0 Steady State Couldown for 16, 24, 32 and 48 EFPYs 0 No instrumentation margins for possible error in pressure and temperature. Please call us, if you need any additional infr-mation. ~ r ( . N. K. Ray I ( Structural M erials & Reliability lechnology. 90 eau ? I 12 gr vp'r-t- -N-T' Tw#-Fr we w -c--e+r w +-ce'-e .r-www e r'whe-rmes-y-w--- v-n=#---'ewr* v'e-ev -,-4,mnL-v-r-m1- -&crmm-Ny awwww-----y op go- =

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Q2.S Jea P' Overoressure-Setnqint Correlation 17

r W N-QD-da W Overores'ure Setooint Correlation s ~ .The analytical correlation between reactor coolant system overpressure and i LTOPS setpoint pressure is derived from a PC based fitting routine, with the assumption of a linear relationship: Setpoint Pressure - ao + a3(Max. Overpressure) The output from the fitting routine is shown below: TERM COEFFICIENT 8 -1.051334000E+02 1 ~ 1.048214000E+00 WHAT NEXT ? 2 X-ACTUAL Y-ACTUAL Y-CALC DIFFERENCE PCT DIFF. 4.8200E+02 4.3000E+02 4.0011E+02 -1.0547E-01 -2.6360E-02 5.7700E+02 5.0000E+02 -4.9969E+02 3.1427E-01 6.2894E 02 6.7300E+02 6.0000E+02 -6.0031E+02 -3.1427E-01 -5.2351E-02 7.6000E+02 7.0000E+02 6.9989E+02 1.0547E-01 1.5069E-02 STD ERROR OF ESTIMATE FOR Y r.3314956 'e e 18

_y \\U t Westinghouse El.ergy Systems g3g Electric Corporation DLW-91-163 June 18, 1991 Mr. K. E. Halliday, Manager Nuclear Engineering PS-DLW-0207 Duquesne Light Company DLCo P0 0-097993 Beaver Valley Power Station P. O. Box 321 Ref: 1) DLW-91-155, 6/06/91 Shippingport, PA 15077

2) DLW-90-528, 1/23/90

Dear Mr. Halliday:

DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNITS 1 AND 2 Steam Generator Tube Plugging Analysis Program LTOPS SetD9jat l Section 4.2 of WCAP-12966 (Duquesne Light Company Beaver Valley Power Station Units 1 and 2, 20 Percent Steam Generator Tube Plugging Analysis Program Engineering and Licensing Report), transmitted via Reference 1 includes an evaluation of the effect of 20% steam generator tube plugging on the Cold Overpressure Transients and the Low Temperature Over;ressure Prevention System (LTOPS) setpoints. The conclusion of the evaluation is that the current setpoints are nonconservative and that the required setpoint change to remove the nonconservatism is a less than 4 psi reduction in the setpoint. The evaluation also concludes that the nonconservatism is small and exceeding the Appendix G limit by this amount will have no impact on the probability of brittle vessel fracture. At the Offsite Review Committee meeting on June 11, 1991, Duquesne Light Company (DLCo) informed Westinghouse that it was their intention to reduce the LTOPS setpoint for Unit I by 4 psi to accommodate the nonconservatism identified in WCAP-12966. DLCo stated that the present Unit 1 LTOPS Technical Specification 3.4.9.3 requires a setpoint of 350 psig and is i applicable for up to 9.5 effective full power years (EFPYs). Via l Reference 2 Westinghouse pt-ovided revised LTOPS setpoints for Units 1 and 2 based on setpoint calculation methodology that does not include instru=cnt unesttaintics. Based on the information provided in Reference

2. DLCo has prepared and submitted to the NRC for approval a technical st.ecification revision to increase the Unit 1 LTOPS setpoint to 444 psig for up to 9.5 EFPYs. Also, DLCo is presently preparing an additional revision to the subject Unit 1 technical specification to incorporate new Appendix G limit curves for up to 16 EFPYs and to revise the LTOPS setpoint to 436 psig consistent with the new Appendix G limits.

Based on the results contained in WCAP-12966 DLCo has stated that they will reduce the 444 psig and 436 psig setpoints in the two proposed technical specification revisions by 4 psi to accommodate the Cold Overpressure

l 4 C' j Mr. K. E. Halliday DLW-91-163 June 18, 1991 Transients associated with 20% tube plugging. Once these changes are approved by the NRC and implemented by DLCo, the technical specification setpoint will prevent low temperature overpressure for the 20% tube plugging condition. Until NRC approval and DLCo implementation of the revisions to the subject technical specification, the installed LTOPS setpoint will not include the 4 p:,i reduction but will continue to include the margin associated with the old calculation methodology which includes instrument uncertainties. Based on this LTOPS setpoint implementation status for Unit 1 DLCo requested Westinghouse to rey;ew the results contained in WCAP-12966 and provide a recomendation on whether the present setpoint of 350 psi should be reduced by 4 psi. Westinghouse has reviewed this item as requested and has confirmed that the present LTOPS setpoint of 350 psig does not need to be reduced by 4 psi. The technical basis for this conclusion is that the margin in the present setpoint of 350 psi is greater than the 4 psi nonconservatism that results from the change in cold overpressure transients due to 20% tube plugging. Consequently, the existing Unit 1 LTOPS setpoint is conservative for tube plugging levels up to the 20% limit. With respect to the proposed technical specification revisions, it is acceptable to reduce the setpoints provided in Reference 2 oy 4 psi to accommodate the 20% tube plugging condition, resulting in Ur.it 1 setpoints of 440 psig for up to 9.5 EFPYs and 432 psig for up to 16 EFPYs. Please advise if there are any questions or coassents on this information or if we may be of additional assistance. Very truly yours, l J. N. Steinmetz, Manager l Central Region Customer Projects Department cc: NERU Records K. Troxler G. Kammerdeiner R. Ireland S. Sovick -}}