ML20238D948
| ML20238D948 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 09/09/1987 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Kerr W Advisory Committee on Reactor Safeguards |
| References | |
| NUDOCS 8709140013 | |
| Download: ML20238D948 (2) | |
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Distribution: DWigginton SEP 0 31987. Docket Files NRC PDR GHolahan Local PDR JHeltemes, AEOD ! Docket No. 50-483 EDO Reading TAlexion l PDIII-3 r/f PKrcutzer TRehm J$niezek FMiraglia MEMORANDUM FOR: William Kerr, Chairman Advisory Committee on Reactor Safeguards FROM: Victor Stello, Jr. Executive Director for Operations
SUBJECT:
PROPOSED j'0WER INCREASE FOR CALLAWAY PLANT Union Electric Company (the licensee) filed an application for a license amendment dated March 31, 1987 (as revised by letter dated April 21,1967) to operate the Callaway Plant at an increased power level from 3411 to 3565 MWt. The staff has begun its review of the proposed amendment. Enclosed is a copy of the application to provide the committee , with the opportunity to review the application. If the connittee does elect to review the application, it is requested that the ACRS report be provided by October 15, 1987. Original Signed By: James M. Taylor ij{)VictorStello,Jr. Executive Director for Operations
Enclosure:
As dtated
Contact:
T. Alexion X24730 8709140013 870909 PDR ADOCK 05000403 P PDR ~ Office: LA/PDIII-3 PM/PDIII-3 PD/PDIII-3 A > AEf7dLy. Surname: PKreutzer TAlexion/tg DWigginton GHolahan FMi 08/gagl f Date: 06/ /87 OW/ 7 06b3/87 Og/87 l Office: p Q P. NR,' NO ED N NR J n'ezek NRR Surname: 0 r' ilhd TMu le) VSt llo 0 Date:
/g/87 Cj/}/87 /j/87 [/87
3,, ~ y , l ., Distribution: Docket Files DWigginton NRC PDR GHolahan Local PDR JHeltemcs, AE0D Docket No. 50-483 ED0 Reading TAlexion m " PDIII-3 r/f PKreutzer TRehm j' JSniezek FMiraglia MEMORANDUM FOR: William Kerr, Chairman Advisory Committee on Reactor Safeguards FROM: Victor Stello, Jr. L. Executive Director for Operations ;
SUBJECT:
PROPOSED POWER INCREASE FOR CA.LLAWAY PLANT I Union Electric Comparf (the licensee) filed an application 1 for a license amendmeX dated March 31,1987(asrevisedby -) letter dated April 21, 7) to operate the Call way Plant at i an increased power level f 3411 to 3565 MWt. The staff has begun its review of-the p osed amendment Enclosed is a copy of the applicatio to pr ide the committee with the opportunity to review the app . ion., If the committee does elect to review the application, it s quested that .I the ACRS report be provided by September 1, I s Victor Stello, Jr. Executive Director ,. for Operations j
Enclosure:
As Stated i 1
Contact:
T. Alexion X24730
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1 4 4 ll /3. Office: LA/PDIII-3 Pp/f1 J- PD/PD1 '3 AD/DRSP Surname: RK Tale >tg DWiaginton GHolahan Date: 08/[a'ctzer y/87 06/jl/87 06/@87 06/ /87 Office: D/DRSP NRR/DONRR ED0 Surname: DCrutchfield TMurley VStello Date: 06/ /87 06/ /87 06/ /87 1 -
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Donald F. SchneII Vee President U.S.LNuclear Regulatory Commission ATTN- Document Control Desk Washing ton, ; D.C. 20555. Gentiemen:- ULNRC-1471. CALLAWAY PLANT " DOCKET NO. S0-483-CALT AWAY- PLANT UPRATING SUBMITTAL Union Electric herewith transmits one original and.one conformed copy of an application'for amendment to Facility Operating License No. NPF-30 for Callaway Plant. B is amendment re^ quest revises Technical Specifications to support a callaway Plant uprating.to the 3565 MWt core power-level. W is will. benefit the general public and customers of
. Union Electric by increasing..the'usefulness of Callaway Plant as
'a safe.and dependable source of power for the. region.- We enclosure to' this letter provides a" listing of. all Attachments.
In summary, the Attachments provide the.'results of: engineering and licencing reviews'that support the'Callaway uprating. These T reviews verify the capability of the plant to operate at the
! increased power level without modifications to existing systems, .
I. structures, or components. Any revisions to the FSAR that are lV required due to the _ uprating will be incorporated into the FSAR
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during the annual update following implementation of the uprating.
' The characteristics of the thermal hydraulic flow anomaly, first discovered at Callaway, are not impacted by the uprating.
Based on information provided to'the NRC in Westinghouse letter NS-NRC-87-3202, dated February 13, 1987, it is expected that
- .- suf ficient margins continue to exist to insure that operation of callaway Plant remains in compliance with all applicable regulatory and industry codes / standards applicable to Callaway at the uprated conditions with the flow anomaly present.
The steam generator tube rupture (SGTR) analysis, which was required by License Condition 2.C.(ll), has been submitted and is the subject of ongoing NRC review. Union Electric's responses to additional NRC questions concerning the SGTR analysis are scheduled for submittal by May 15, 1987. 870406049W B70331 cPDR ADOCK 05000483
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This request has been reviewed and approved by the CallawayIt Onsite Review Committee and the Nuclear Safety Review Board. has been determined that this request does not involve any unreviewed safety questions as defined in 10CFR50.59 nor a significant . hazard consideration as determined by the three factor test per 10CFR50.92. Union Electric proposes October 15, 1987 as the effective date for implementation of the uprating, subject to NRC approval. Enclosed is a check for the $150 application fee required by 10CFR170.21. Very truly yours, Donald F. Schnell DJh'/ mat Enclosure Attachments
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. STATE OF MISSOURI )
)' ~ SS CITY OF ST. LOUIS')-
Donald'F. Schnell, of lawful ege, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of: Union Electric Company; that he has read the foregoing document and -l knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his 3 knowledge, information and belief. Donald F. Schnell Vice President Nuclear SUBSCRIBED and sworn to before me this 3/ 4 day of , 1987 BA:BARA J. FFAFF vY NOTARY P.iUC, STATE OF WISSOURI Ki COMMISSN EXPlR[$ APRit 22, 1989 3 ST. LOUIS COUNTY
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i-Ecc ': Gerald Charnoff, Esq. Shaw, Pittman,aPotts & Trowbridge 2300 N. Street, N.W. Washington, D.C. 20037 J. O. Cermak-CFA Inc. 4 Professional Drive (Suite 110) Gaithersburg, MD 20079 W. L. Forney Division of Projects and! Resident Programs, Chief, Section 1A U.S.. Nuclear Regulatory Commission Region III-799 Roosevelt Road Glen Ellyn, Illinois 60137
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Bruce Little Callaway Resident Office ! j
;: U.S.J Nuclear Regulatory Commission RRll Steedman, Missouri 65077
( Tom Alexion (2) Of fice of -Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 316 7920 Norfolk Avenue ! Bethesda, MD 20014
- Ron Kucera, Deputy Director i' Department of Natural Resources ,
P.O. Box 176 Jefferson City, MO 65102 Manager , Electric Department Missouri Public Service Commission f P. O. Box 360 Jefferson City, Mo 65102 l 1 1 4
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). > D.- F. Schnell
~J . E,. Birk .
1 J. F. McLaughlin A. P. - Neuhalfen '
' R.E J . Schukai M. A. Stiller S. E. Miltenberger , ,
D.'E.-Shain H. Wuertenbaecher , D.'.W. Capone A. .C. Passwater W. R. Campbell R. P.,Wendling D. E. Shafer D. J. Walker
-O. .Maynard.(WCNOC)
N. P. Goel-(Bechtel) G56.37 (CA-460) > Compliance (J . E. ' Davi s )
'NSRE.(Sandra Auston)- '
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' <D ; o ULNRC-1471 March-31, 1987 j CALLAWAY UPRATING SUBMITTAL LIST OF ATTACHMENTS Attachment 1 Safety. Evaluation !
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' Attachment.2 10perating License / Technical .
. Specifications Revised Pages ,
l Attachment 3 Environmental Evaluation Attachment 4 .Significant Hazards Evaluation-
. Attachment 5 Appendice s v Appendix A -~NSSS Uprating Licensing Report Appendix B - Section I BOP Uprating - Licensing Report . .Bechtel Review 1
Appendix B - Section II BOP Uprating - Licensing Report Union Electric Review 4 l ~ _ - - - - - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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-ULNRC-1471 March 31,.'1987' 8
criteria are met. Refer to ULNRC-1470 dated .3/31/87 for a detailed .6einription of the analyses. IThe"NON-LOCA.analysesofrecordwereevaluatedforimpact Certain cases were
' due/to V-5 fuel (ULNRC-1470 dated 3/31/87).
reanalyzed in order.to support the V-5' fuel. These reanalyzed
" .NON-LOCA cases were performed at an NSSS power of 3579 MWt and the assumed. parameters are consistent with the uprating.
Therefore,.the result,s are valid for.the Callaway uprating. Results show that all applicable safety criteria are yet. Refer
.to ULNRC-1470 ' dated 3/31/87 for a detailed description 4 cf1 these, analyses.
j The licensing basis supported by this report continues to be' 10% maximum steam generator tube plugging. Per ULNRC-1470 dated A 3/31/87, 15% tube plugging was assumed'for all LOCA cases and for 3
-the reanalyzed NON-LOCA cases. However, the NON-lDCA cases which "'*?
were evaluated and were not impacted by V-5 fuel t(and
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r consequently'not reenalyzed) only assume 10% tube plugging. ,4
\
The radiological source terms were: evaluated as part of the Cycle 3 reload analysis at the uprated power level. The source terms are based on 102% of 3565 MWt core power. Refer to ULNRC-1470 dated 3/31/87 for a detailed description of the radiological analysis. , , Four additional MSLB cases were analyzed to determine the i impact of the uprating. Cases 1,2,3, and 16 of FSAR Table
, 6.2.1-56 (representing 102% of 3425 MWt) were reanalyzed at 102%
.e of 3579 MWt.(raising the spectrum of cases from 16 to 20).
The f g four.addi,blotwil P/T curves are bounded by the ez3 sting Equipment i Qualificati k (EO) envelope with o m exception.4 The temperature profile of reanalyzad Case 16 exceeds the EQ ' envelope at one point: howe 4ar, the overall case is less limiting than the EQ
. envelope. %&refore, no impact on equipment qualification will occur Mue to the uprating. Refer'to Attachment 5, Appendix B, Sectica V for further discussion.
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j
e e RELEASES DURING NORMAL PLANT OPERATION The Callaway Plant releases and offsite doses were previously evaluated by the Staf f relative to 10 CFR 50 Appendix I for an uprated reactor power level of 3565 Mwt' in Section 11.1 of NUREG-0830,.SER, October 1981. In the SER, the Staff concluded that the liquid- and gaseous- waste-treatment systems , will be capable of meeting the ALARA levels and the requirements of 10 CFR 50.34a and Appendix I to 10 CFR 50. lI _ .;3 ,. r
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SUMMARY
OF . TECHNICAL. SPECIFICATIONS CHANGES t
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f# Table 2 presents a summary list of the:Technidal /
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,-o- 5 specification Changes due to the..uprating/.to a licensed core ~ power.of 3565 MWt (NSSS power of 3579 MWt)'.
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.. N CON,'.'LU SION S 1/ /
/ This review h'as demonstrated that the Callaway Plant is capablen in its present design' configuration, of operating safely and rejiably at the ' proposed power rating and remains in compliance with the design criteria and safety limits specified in the FSAR, provided the pla.it' is operated in accordance with the proposed Technical Specification changes. The review has verified the following: ,
- 1. The probability of an accident previously evaluated in the FSAR will not be increasad.
2( The consequences of an accident previously evaluated in j the FSAR will not be increaaed.
- 3. The possibility of an eccident which is different than any already evaluated'in the FSAR will not be created.
.- 4. The probability of a. malfunction of NSSS or BOP ' equipment important to safety, previously evaluated in the FSAR, will not'bo increased.
- 5. The consequences of a malfunction of NSSS or BOP l equipment important to safety, previously evaluated in the FSAR, will not be increased.
,. g,c t.' 6 ', The possibility of a malfunction of NSSS or BOP equipment important to safety, dif ferent from any
! already evaluated in the FSAR, is not created by operation at the uprated power.
V The margin of safety as defined in the bases to any t 7. [ technical specification will not be reduced by operation - at th6 uprated power. [- h There fere, i t 'has been concluded that operation of the Callaway Plant,,at'the increased power rating does not reduce any
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cafety macgins, and'does not involve an unreviewed safety question as defined by 10 CFR 50.59. l
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y i 9 .. l 3 Table 1 CALLAWAY UPRATING SYSTEM REVIEW LIST Class Appendix System. (Note 1) Description (Note 2) AB .O Main Steam' B AC 0 Main Turbine' B AD N' Condensate B AE' O Feedwater B AF N FW Htr. Extrac., Drns., Vents, B AK- N' Condensate Demineralized B AL 0 Auxiliary Feedwater B
^ AM N Raw Water Supply X
-AN S Demin. Water Stor. & Transfer B AP- O Condensate Storage & Transfer B AO N Cond. & FW Chemical Add. X BB O- Reactor Coolant .A BG O Chemical & Volume Control A BL O Reactor Makeup Water B ,
BM O Steam Generator. Blowdown B { BN O Borated Refueling Water Stor. B CA N Steam Seals B I CB N Main Turbine Lube Oil B CC .N Generator H2 & CO2 . B , CD N Generator Seal Oil B CE N Stator Cooling' Water B
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CF N Lube Oil' Stor . , . Tran . , Purif. B CG N Condensor Air Removal B l CH N Main Turbine Control Oil B ! l DA N Circulating' Water B DB N Cooling Tower Makeup & Blowdown B DD N Circ. Water Chemical Control X DE N Intake & Water Treatment B
. EA N Service Water B Closed Cooling Water B ;
EB N EC O Fuel Pool Cooling & Cleanup B EF 0 Essential Service Water B l EG O Component Cooling Water B EJ O Residual Heat Removal A EM Q Hig~n Pressure Coolant Injection A EN O containment spray B 1 EP O Accumulator Safety Injection A j i l L __-__2_-- _ _ _ _ -
Table 1 (Cont.) FA N Auxiliary Steam Generator X i FB O Auxiliary Steam X FC O Auxiliary Turbines . B FE N Aux. Steam' Chemical. Add. X 4 GA S Plant ~ Heating X GB N Central Chilled Water B GC N Service & Stores. Bldg. HVAC X GD O ESW Pumphouse HVAC X GE O Turbine Building HVAC X GF Q Misc. Building HVAC X GG O Fuel Building ITVAC - B GH S Radwaste Building HVAC X i GK Q Control Building HVAC X Auxiliary Building HVAC B GL O oGM O Diesel Generator Building HVAC X GN O Containment Cooling B GP O Containment ILRT X GR N Containment Atmosphere Control B. GS O Containment Hydrogen Control A i Containment Purge B GT O' HA S Gaseous Radwaste B Licuid Radwaste B HB .O HC S Solid Radwaste B '1 HD Q Decontamination X S Boron Recycle B HE ! S Secondary Liquid Waste B HF JA N Aux. Oil Stor. & Transfer X JE O Emergency Fuel Oil X JH N Auxiliary Gas X ! KA O Compressed Air X KB O Breathing Air X KC O Fire Protection X KD N Domestic Water X KE O Fuel Hand., Stor., & RV Serv. X Cranes, Hoists, & Elevators X KF S KH S Service Gas X Standby Diesel Engine X KJ O KS N Bulk Cr.emical Storage X Sanitary Drainage X LA S Roof Drains X LB S LC N Yard Drains X LD N Chemical & Detergent Waste X i LE O Oily Waste X J ' LF 0 Floor & Equipment Drains X 1 y i
o , a Table 1 (Cont.) MA N Main Generator- B MB- N Excitation & Voltage Regulation B MD N EHV Switchyard Bus B ME .N EHV' Switchyard 125 V DC B MF N EHV Switchyard Lighting X MH N Outgoing EHV Lines B MR N Startup Transformer B NB. O 4.16 KV AC (lE) B NE O Standby Generator . B NF 0 Load Shed. & Emer. Load Seq. X NG- 0 480 V AC (lE) B NK O 125 V DC (lE) . B NN O. Instrument AC (IE) B PA O 13.8 KV.AC B PB N 4.16 KV AC B PG O 480 V AC B PJ N 250 V DC B PK O 125 V DC B PN O Instrument AC B PO N Uninterruptible AC B OA N Normal Lighting X OB O Standby Lighting AC X OD S Emergency Lighting DC X OE N Telephone System X OF S Public Address X ..}- OG N Grounding , X , s OH N Cathodic Protection X
- . OJ N Freeze Protection X ON N Misc. Equipment X OT N Permanent Road Lighting X RD N Meteorological Instrumentation X RJ N BOP Computer B RK N Plant Annunicator X RL 0 Main Control Board X RM N Process Sampling B RP O Misc. Control Panels X RR O Safety Assessment System X SA O BOP ESFAS X t SB O Reactor Protection A SC O Reactor Instrumentation A SD N Area Radiation Monitoring X SE O Ex-Core Meutron Monitoring A SF N Reactor Control A SG N Seismic Instrumentation X SJ O Nuclear Sampling B 4
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Table 1 (Cont.) Plant Security X SK N-SP Q Process Radiation Monitoring X Loose Parts Monitoring X SO N-Incore Neutron Monitoring A SR N ST- N Emer. Resp. Facility In'fo. X Radioactive Release Info. .X SZ N Notes:
- 1. Class: 0 =. system is-safety related or has. safety related components.
S = System is special scope (FP, DA, 2/1) or has special scope components, but has no safety related function or components. N = System is non-0, non-special scope, ar,d has no safety related or special scope components.
- 2. Appendix: A = Appendix A of Attachment 5 (NSSS Licensing Report).
B = Appendix B of Attachment 5 (BOP Licensing Report). X = Review not required; system operation not dependent on thermal power level.
R eb P r s U ile uls E qi y d d H ewl e e T r a t e t Rn. a h a O reare r p t r T eW - p gOtw u m u S nPno ep o E o e r e G lLd h f h N Aie t t A oMcr t H nRco s l s C Eac t u t sH c s c N iTes e e e O - hi l r l I nDth f f T oE t e s e A N iTd r t . r C O tant ie I I iRaa e g mr e g F T n io I A ie,d n lc n C C fhte a - a E I etWm h wd h P S F I T d sc5f e Mr o c se ee nt a c. se L S i n6 r ir er ir . A U hi5e ho hp ho C J I
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2 N H L L . e C A A s hs l E M Mse t te b T R . Rnu i it T a T Es Eol m wo Hn Hia i nd N To Ttv L R n A i i Eea L P Nt Nnd y Ws Gi Gie t Oi T
.., I n If s e Pv Y Si Sei. f ea A Ef EdvR a LRt W De D eE S A- l
= A E d dRW M e. e - L G f ,. f e O R .ds L N ore or P r ER t A A e. o HETn TWOi C H nb nb1L C C om A O o R iu io um 4. M r NPhp O F tn tn1R o G tt F O i- i - E t ILie ne nehH c Saws N N ir i r gT a EM O O f f u e DRdT I I ed edoD R Ee T T dn dnrE dHta A P a ahT d etat C I e e tA e c il I R tR tR R s aDce F C eE eE1 i lE o d I S lW lW1 r v pT s - T E eO eO .o e eAsP S D DP DP1f R RRaO U J 1 D 2 N . A N 2 O e Y I x - r1 e R T e 01 u- l A C d 14 g1 b M E n . i . a M S I 11 F2 T U S E I - ,0 G I 27 2 71 A - - - - - P I 11 2 22 i
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CALLAWAY UPRATING SUBMITTAL P ATTACHMENT 2 OPERATING LICENSE AND TECHNICAL SPECIFICATIONS .j. REVISED PAGES .p {,. . I l
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FA CZLTrf OfMRr.rNG tr ccy.rg Nff 30 (4) UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or phys-ical. fom, for sample analysis or instrument calibration or asso-ciated with radioactive apparatus or components; and (5) UE, pursuant to the Act and 10 CFR-Parts'30,' 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. ( C. This license shall be deemed to contain and is subject to the conditions specified in the Comission's regulations set fe,rth in 10 CFR Chapter I . and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Comission now or hereafter in effect; and is subject to the additional conditions
- specified o- incorporated below:
S (1) Maximum power Level 3C45 UE is authorized to operate' facility 'at reactor core power levels not in excess of Emegawatts thermal (100% power) in accordance with the conditions specified herein and in Attach-ment 1 to this license. The preoperational tests, startup tests and other items identified in Attachment 1 to this license shall be comoleted as specified. Attachment 1 is hereby incorporated F into this license. . (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A a#nd the Environ-mental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated' into this license. UE shall operate the facility in accordance with the Technical Spec- . ifications and the Environmental Protection Plan; (3) Environmental Ouelification (Section 3.11 SSER #3)* Nove *** (a) Prior to L .mfwM..,v1985. UE shall environmentally qualify all electrical equipment accordingly to the provisions of 10 CFR 50.49. , (b) Prior to restart following the first refueling outage, UE shall have qualif,ied the reactor vessel level instrumentation system high volume sensor.
'The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed. .
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8704060 d 070331 PDR' ADOCK 05000483
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INDEX DEFINITIONS PACE 5ECTION ,
- 1. 0 DEFINITIONS 1.1 ACTI0N............................................................, 1-1 1.2 ACTUATION LOGIC TEST....... ....................................... 1-1 1.3 AN A LO G CHA NN E L O P E RAT I O N A L T E ST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.4 AXIAL FLUX DIFFERENCE.............................................. 1-1
- 1. 5 CHAN N E L CA LI B RATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 1-1
- '1. 6 . CHANNEL CHECK...................................................... 1 - 1.7 CONTAIMMINT INTEGRITY.............................................. 1-2 1.8 C O NT R O L LE D LEA KA G E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2
- 1. 9 C O RE A LT E RAT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 vs0/tEs sus TnEsniceoc e. ./. . /. . . /. . ./ . . ./. . . s. . s. . ./. . ./. . g ;-; ;
1-2
- 1. X /0 DO S E E QUI V A LE NT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3
- 1. X// E- AVE RAGE DISI NTEGRATION ENE RGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3
- 1. W2 EN.3INEERED SAFETY FEATURES RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3 1.)/./J F R E QU E NC Y N0T ATI ON. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.J4TIDENTIFIEDLEAKAGE.................................................
- 1. )y.fMAS T E R RE LAY TE 5T . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . .
1-3 ! 1-3 ;
- 1. M/MEMB E R ( 5 ) 0 F THE P UB LI C. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '
1-4 :.
- 1. u/ @F F SIT E 00sE CALCU LATION MANU AL. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-4 !
- 1. )fff0 P E R AB LE - 0 P ERAS I LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . n . . . . . . . . . . . . . . .
- 1. g/( O P E RATI ONA L MO D E - MO D E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-4 1-4 i
- 1. y& HY S I CS T E S T 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i 1-< :
1.M J.fPRES$URE BOUNDARYLEAKAGE.......................................... I 1-4 j 1.r u ROCE35 C0 m ot eR0 m M............................................ 1-< j 1.y.s PUR G E - r unG I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5
- 1. nt@U AD RANT r0VE R TI LT RAT 2 0. . . . . . . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . .
1-r 1.pg R AT Eo TxE RnA L P0vE R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.p)/,RE ACTOR TRIP SYSTEM RE SPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 l l 1. p$ E P 0 mB L E EVE NT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5 , l 1. ;r;,rs NuTo0wN xxRc ! N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I Amen ce d N:- 15 ) f CALLAWAY - UNIT 1 1 1 l
l .. .. ..
~
l INDEX l DEFINITIONS SECTION
- _PAGE DEFINIT 1DNS (Continued) 1.,701$51TE 80VNOARY............................................y........ 1-5 1.Jiy tE LAVE RE LAY TE ST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 l 1-5
- 1. )f)/ S O LI D I F I C AT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1.)5'>5OURCE J CHECK...................................................... 1-5 1-6 L 1.)(155TAGGEREDTEST8 ASIS.............................................. 1 .
- 1. )6{ f4hE RXA L P 0VE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1.)(IfTRI P A;TUATING D EVI CE OPE RATIONAL TE57. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1,7344UN I D E NTI F I E D LEA KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6
- 1. Jh$ p uN RE 5 7 R I C T ED AR EA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 l_
1.79/ VENTILATION EXHAUST TREATMENT SY5 TEN.............................. 1-6 1.9034VENTINS........................................................... 1-6 1,,f14#WA STE GAS H O LDU P S Y5 TEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 T ASLI 1.1 FREQUENCY NOTATION........................................... 1-8
) TABLE 1.2 OPERATIONAL MODE 5............................................ 1-9 t
1 i l 1 II Amendment No.15 CALLAWAY - UNIT 1 f
DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAIW.ERT INTEGRITY shall exist when: -
- a. All penetrations required to be closed dcring utdent conditions are either:
- 1) Capable of being closed by an OPERABLE containment automatic isciation valve system, or
- 2) Closed by manual valves, blind flanges, er deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification
- 3. 6.1. 3, , ,
- d. The containment leakage rates are within the limits of Specification
- 3. 6.1.2, and .
- e. The sealing mechanism associated with each pe etration (e.g., welds, bellows, or 0-rings) is OPERAELE.
4 CONTROLLED LEAKAGE O j' 1.8 CORTROLLED LEAKAGE shall be that seal water flow f em the reactor coolant { pump seals. ' CORE ALTERATION
- 1. 9 CORE ALTERATION shall be the covement or manipulation of any component i within the reacter vessel with the vessel head removed and fuel in the vessel. -
l Suspension of CORE ALTERATION shall not preclude completion of sovament of a component to a safe conservative pcsition. DES}GN THEP.AL POVEV 10 D !GN THE AL POWER hall be design res. .r core est tr sfe ate the rea ter coola of 3555 dt. DOSE EQUIVALENT I-131 )
/0 1.R DOSE EQUIVALEhT I-131 shall be that concentration of I-131 (sierocurie/gra-)
1 l l which alone would produce the same thyroid dose as the cuantity and isotopic l sixture of I-131,1-132. I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test ! Reactor Sites." i r l CALLAWAY - UNIT I 1-2 Ament. ment No. U
j l der 1N!TIONS T - avEnacE etsikTEcRATION ENEecY
- 1. I shall be the average (we'ighted in proportion to the concentration of l each radionuclides in the reactor coolant at the time of sampling) of the sum of the average beta and gama energies per disintegration (in MeV) for isotopes, othe- than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
ENGINEERE.D SAFETr FEATURES RESPONSE TIME l 2.
- 2. ~ The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time l l 1 terval from when the sonitored parameter exceeds'its ESF Actuation Setpoint at the channel sensor until the ESF equipeent is capable of performing its safety function (i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. .
FRE0VENCY NOTATION
'~
(3 ' 1.J4 The FREQUEN0Y NOTATION specified for the performance of Surveillance [ Requirements shall correspond to the intervals defined in Table 1.1. ,
~
1 DER IrIED LEAKAGE , s
- 1. IDENTIFIED LEAKAGE shall be: l
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as .
' pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere free sources that are both j specifically located and known either not to interfere with the t' operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or ,
b
- c. Reactor Coolant System leakage through a steam generator to the i
Seconda y Coolant Systen. MASTER RELAY TEST
.1. A MASTER RELAY TEST shall be the energization of each master relay and l verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay. j 1
I MEMBER (5) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa- l 1. tionally associated with the plant. This category c'oes not include ecployees of the licensee, its contractors or vendors. Also excluded from this categcry are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recrea-tional, occupational, or other purposes not associated with the plant. 1*3 Amendt,ent No.15 CALLAWAY - UNIT 1 ___-_______,,,,_m
DEFIN!T10N5 OTFSITE 005E CALCULATION MANUAL
- 1. The DFFSITE DOSE CALCULATION MANUAL (ODOM) shall cortain the methodology j and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent sonitoring Alarst/ Trip Setpoints, and in the conduct of the Envircreental ,
Radiological Monitoring Prograai. OPERABLE - OPER8.BILITY
- 1. A system, subsystem, train, component or device shall be OPERABLE or l ,
- have OPERABILITY when it is capable of perforzing its specified function (s), I i
and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are l required for the system, subsystem, train, component, or device to perfors its -
-function (s) are also capable of perfornirg their related support function (s). .
l
. OPERATIONAL MODE - MODE {
- 1. An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l combination of core reactivity condition, power level, and average reactor 7
coolant temperature specified in Table 1.2. , l PHYSICS TESTS 2 PHYSICS TESTS shall be these tests perfonced to seasure the fundamental i j pi' nuclear characteristics of the core and related instrumentation: (1) cescribed in Chapter 14.0 of the FSAR, or (2) authorized under the provisions of l
' 10 CFR 50.59, or (3) otherwise approved by the Comission.
PRESSURE BOUNDARY LEAKAGE i l i 1. PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generater tube ] j 1 skage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. . p
; PROCESS CORTROL PROGRAM The PROCESS CORTROL PROGRAM shall contain the current formula, sa pling. l 2 I analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or sir.ulated wet solid wastes will be accomplished in such a way as to assure . compliance with 10 CFR Part 20,10 CFR Part 71 and Federal and State l
regulations, burial ground requirements, and other requirements governing the disposal of the radioactive waste. l PURGE - PURGING 1 k PURGE or PURGINO shall be any controlled process of discharging air or l g(s fron a confinement to maintain temperature, pressure, htatidity, concentra-tion or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. 4 1*4 42fnC2thi N0' CALLAWAY - UNIT 1
DEFINIT 20NS - QUADRANT POVER TILT RATIO
- 1. QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excere l detector calibrated output to the average of the upper escore setector cali-brated outputs, or the ratio of the maximum lower excere detector calibrated output to the average af the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors .
shall be used for computing the average. RATED THERMAL POVER 1 RATED THERMAL POWER shall be a total core htet transfer rate to the l reacter coolant of Mit n't . , t REACTOR TRIP SYSTEM RESPONSE TIME
~
' 1. The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from j when the eenitorad parameter e.xceedr. its Trip 5etpoint at the channel senser until loss of stationary gripper coil voltage. I
; REDORTABLE EVENT
- 1. A REPORTABLE EVENT shall be any of those conditions s'pecified in l Section 50.73 to 10 CFR Part 50. ,
SN'JTDOVN MAR 3!N , A.
- y !
- ' 1.t1 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l I the reactre is subcritical or would be suberitical from its present condition assueing all full-length red cluster assemblies (shutdown and control) are fully inserted axcept for the single red cluster asserbly of highest reactivity worth which is assumed to be fully withdrawn.
1 5fTE BOUNCARY , j 1 The SITE BOUNCARY shall be that line beyond which the land is neither l owned, nor leased, nor otherwise controllac by the licensee. SLAVE RELAY TEST
- 1. A SLAVE RELAY TEST shall be the energization of each slave relay and l verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices. )
$0L1DIFICATION
- 1. ! SOLIDIFICATION shall be the conversion of wet wastes into a form that I meets shipping and burial ground requirements. I SOURCE CHECK -
)
- 1. A SOUR E CHECK shall be the qualitative assessment of channel response I when the channel sensor is exposed to a source of increased radioactivity.
CALLAWAY - UNIT 1 1-5 Amendment No.15
i DEFINITIONS .
~
STAGGERED TEST SA515 1.)( A STAGGERED TEST BASIS shall, consist of: -
-l
- a. A test schedule for n systems, subsystems, trains, or other de,signated components 'obtained by dividing -the specifies test interval into n equal subintervals, and
- b. The testing of one system, subsystem,' train, or other e,esignated coeconent at the beginning of each subinterval.
THERMAL POVER , 1.)96HERMAL PDWER shall be the total core heat transfer rate to the reacter l coolant.
. TRIP ACTUATING DEVICE OPERATIONAL TEST .
A TRIP ACTUATING DEVICE OPERATIONAL TEST shall' consist of operating the 1.j[ Trip Actuating Device and verifying OPERABILITY of alars, interlock l trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustaent, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy. UNIDEh*TIFIED LEAKAGE
~
4
'1.[f UNIDENTIFIED LEAXAGE shall be all leakage which is not IDEhTIFIED LIAKAGE l ; or CONTROLLED LEAKAGE. : ' 3 UNRESTRICTED AREA ; 2.[ An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l access to which is not controlled by the licensee for purposes of protection of l i, '
individuals from exposure to radiation and radioactive saterials, or any area
.within the SITE SOUNDARY used for residential quarters or for industrial, } cosenercial, institutional - and/or recreational purposes. -
I VECRATION EXHAUST TREATMENT SYSTEM i: 1.N A VENTILATION EXHAUST TREATMENT SYSTDi shall be any system' designed and installed to reduce gaseous radiciodine or radioactive material in particulate
]
form in effluents by passing ventilation or vent exhaust gases through charcoa; adsorbers and/or HEPA filters for the purpose of. removing iodines or partie-ulates from the gaseous exhaust stream prior to the M1 ease to the environment. Such a system is not considered to have any effect on noble gas affluents. Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING 1.k VENTING shall be any contre 11ed process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not pro-vided or moutred during VENTING. Vent, used in system names, does not imply a VEh71NG process.
]
I l 1 CALLAWAY - UNIT 1 1-6 Amenament No. U j i i I
( - p]Ef)N1710NS . WASTE CAS NOLDUP SYSTEM 1 A WASTE CAS HOLDUP SYSTE.M shall be any system designed and installed to l reduce radioactive gaseous effluents by collecting Reactor Coolant System off-gases from the Reactor Coolant Syster and providing for celay or holdup fer the purpose of reducing the total radioactivity prior to release to the environment. l l l l l l l-l l l l.. 4 9 O 4 i 1-7 Amenceent Nc.15 CALLAWAY - UNIT 1 l 1
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Attachment 3 ULNRC-1471 March 31,.1987 CALLAWAY UPRATING SUBMITTAL ATTACHMENT 3 ENVIRONMENTAL EVALUATION Union Electric has performed an environmental evaluation to support uprating the Callaway Plant from.the currently licensed power level to the core thermal power of 3565 MWt which corresponds to a NSSS power level of 3579 MWt. This change does not provide a significant adverse environmental impact, nor a significant' increase in effluents, nor an adverse land or-cultural resource altering activity. The original licensing i' evaluations for the plant, including the NRC Environmental Evaluations (Ref. NUREG-75/011, 3/75, Section 1.1), were based on a NSSS. thermal power level of 3579 MWt. Therefore, the proposed , uprating remains within the bounds of the original environmental analyses. 9 'Y. c'- F.i l l
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Attachment 4 ULNRC-1471-March 31,'1987 CALLAWAY UPRATING SUBMITTAL ATTACHMENT 4 SIGNIFICANT HAZARDS EVALUATION
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t Attachment-4 ULNRC-1471 .! March 31, 1987 l 4 SIGNIFICANT HAZARDS EVALUATION FOR CALLAWAY PLANT UPRATING TO 3579 MWt This evaluation supports Union Electric Company's license amendment request to increase the thermal output of.the Callaway
-Plant. At present, .Callaway is licensed to operate at a reactor core thermal power level of 3411 MWt, which corresponds to a Nuclear Steam' Supply System (NSSS) thermal power level of 3425 MWt. This amendment requests the necessary License and Technical Specification changes to operate t'he Callaway Plant at a reactor core thermal power level of 3565 MWt, which corresponds to an NSSS thermal power level of 3579 MWt. This represents an increase of 4.5% and is equivalent to.the Engineered Safety Features Design Rating of the Callaway Plant. Union Electric has submitted a license amendment request (ULNEC-1470, dated March 31, 1987) to utilize Westinghouse 17x17 VANTAGE 5 fuel beginning with Callaway Cycle 3 operation. The requested uprating is scheduled to begin concurrently with cycle 3 operation.
In order to implement the uprating, Callaway Plant, Facility Operating License NPF-30, Section 2.C.(1) requires revision to indicate operation is authorized at reactor core power levels not in excess of 3565 megawatts thermal. Technical Specification changes include revision of the definitions of Design Thermal Power (Section 1.10) and Rated Thermal Power (Section 1.26), rescaling of Figure 2.1-1 (Core Safety Limits), and revision of , the notes associated with OT Delta T and OP Delta T (Table 2.2-1). In support of the requested uprating, the NSSS system and component designs were analyzed to verify continued compliance with licensing criteria and standards currently required by the Callaway operating license and the f unctional requirements specified in the FSAR for operation at 3579 MWt. The review was conducted in accordance with the methodology documented in Westinghouse topical report WCAP-10263, "A Review Plan for Upra*,ing the Licensed Power of a Pressurized Water Reactor Power Plant". This methodology has been referenced in connection with the approval of similar license amendments for other facilities. WCAP-10263 methodologies were also utilized in the review of the interfaces between the NSSS and the Balance of Plant (BOP) systems and components. Reviews of the BOP systems and compr.nents were conducted and verify their capability to meet the requirements for operation at 3579 MWt. Review of the turbine generator is based on the Valves Wide Open heat balance which represents an NSSS power of 3562 MWt. The ability of the turbine generator to operate up to 3579 MWt will be determined by careful
^
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, (
monitoring of plant performance parameters. These:revie[ws 3
.demonstrateythat the Callaway. Plant is capable, in the present design 1 configuration,fof operating at the uprated. conditions without 'v$61ating any of the design or safety limits specified in the Callahay FSAR and Facility Operating License NPF-30.
Ih conjunction with the use of Optimized. Fuel Assemblies (OFA) 'during Callaway Cycle 2 operation, accident analyses were c performed at the uprated power conditions'. These-were documented in ULNRC-1207, dated '11/15/85 and ULNRC-1247, dated 1/28/86. fin conjunction with the change to VANTAGE 5 fuel for Cycle 3, .;
.certain of the. accident analyses were redone. These analyses ,
n were also based on the.uprated power condition and are documerted in ULNRC-1470, dated March 31, 1987. The analyses that. wore'not'
-redone are those that are still bounded by the OFA analysess performed for Cycle 2. The results of the accident analyses demonstrate that all safety limits are met at the uprated power conditions.
Radiological dose consequences pertaining to the auprdted ~- power conditions have been completed and documented as pant of , I the license' amendment in support of Callaway Cycle 3 for the utilization of VANTAGE 5 fuel, which are docutented in ULNRC-1470 dated March 31, 1987. The evaluation is also applicable to this amendment request. The environmental impact of operating Callaway Plant at_the uprated condition was previously evaluated in the Callaway Environmental Report as a bounding assumption. This has been approved by NRC in NUREG-75/011, Section 1.1 da'ted 3/75., e) This change does not involve a significant increase in
; the probability or consequences of an accident previous 2y evaluated. This is based on the '
environmental evaluation being previously approved at the uprated conditions and the radiological evaluation which is presented in the Callaway Cycle 3 amendient ' request for the utilization of VANTAGE 5 fuel. , .c b) This change does notL ezoate the possibility of a new or oifferent kind of accident from any accident previously evaluated. Thin is based on system and component reviews which verified their capability to operate at 6 the uprated conditions.f
~
c) This change does not inyolve a significant reduction in a margin of safety. Accident analyses were performed at the uprated conditionn and demonstrate the DNB . design basis remains unchanged, the RCS pressure limit of 2700 psig is not exceeded, and the LOCA results L renu 2n well below the regulatory limits given in 10CFR50.46. Based oh the above discussions, the amendment request does not involve e significant increase in the probability or g 1 L
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.j- .
- 3_
consequences of an accident previously evaluated; does not create f.he possibility of a new'or different kind of accident from any
. accident'previously evaluated; does not involve a reduction'in the required margin of safety. Based on the foregoing, the requested amendment does not present a significant hazard.
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CALLAWAY_UPRATING SUBMITTAL- l' , e I g Y )' ATTACHMENT 5: APPENDICES- p .e APPENDIX A
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- UNION ELECTRIC CALLAWAY PLANT NSSS UPRATING - ~3579 MWt LICENSING REPORT TABLE OF CONTENTS f.ag OBJECTIVES i i
SUMMARY
. ii L 1.0 - INTRODUCTION 1-1 -
' j F
l 2.0 CDKPARISDN OF PARAMETERS 2-1 l-e l 3.0- ACCIDENT ANALYSES 3-1 3-1 [ . 3.1 Wen-LOCA Events 3.2 LOCA Events 3-1 3.3 Hot Leg Switchover . 3-1 3.4 Hydraulic Forces
- 3-2 .
~
4.0 WSSS COMPONENTS IMPACT 4-1 4.1 Basis for Evaluation 4-1 4.2 Equipment Reviews 4-1 4.2.1 Reactor Vessel 4-1 Reactor Internals 4-2 { 4.2.2 t 4.2.3 Reactor Coolant Pumps 4-3 4.2.4 Control Rod Drive Mechanisms 4-3
~
4.2.5 Reactor Coolant Piping 4-4 4.2.6 Pressurizer 4-4 4.2.7. Steam Generators 4-4 5970s.1d/031887
UNION ELECTRIC CALLAWAY PLANT l- NSSS UPRATING - 3579 MWt t! CENSING REPORT TABLE OF CONTENTS (Continued) I Pace 4.2.8- RCS Component Supports 4-4 4.2.9 Nuclear Fuel 4-5 4.2.10 . Auxiliary Systems Components 4-5 4.3 Conclusions 4-5 5.0 NSSS SYSTEMS REVIEW 5-1 5.1 Basis of Evaluation 5-1 5.2 Systems Evaluation 5-1 5.2.1 ' Fluid Systems 5-1 5.2.2 Control Systens 5-3 5.2. 3 Protection Systems 5-3
'5.2.4 AMSAC 5-3 5.3 Conclusions 5-5 -
6.0 NSSS/BDP INTERFACES 6-1 6.1 Introduction 6-1 6.2 Wass and Energy Release Data 6-1 6.3 Auxiliary Feedwater System 6-1 6.4 Radiation Source Terms 6-3 6.5 Component Cooling Water Interface Requirements 6-3 5970s:1d/031887
.- a
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[ , . . L. UNION ELECTRIC CALLAWAY PLANT tc NSSS UPRATING - 3579 MWt LICENSING REPORT LIST OF TABLES
/
Table Title M 2-1 Ce11away Power Capability Parameters - 2-3 5-1 . Systems Reviewed for Uprating Implementation - 5-2 5-2 Callaway Plant Cooldown for 3579 MWt 5 s i 6 System Performana Requirements Reviewed for Uprating Implementation 6-2 I i . 1, 6 1 -I l I e 597ee 16/031687 l
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. UNION ELECTRIC' CALLAWAY PLANT.
- NSSS UPRATING - 3579 MWt LICENSING' REPORT,
, LIST Of FIGURES
- Figure Title Pne, 1 Comparisen of Reactor. Coolant' System' Temperatures 2-4 vs. Percent Rated Load I
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l l 1 5970e id/031887 h____..___ _ _ . _ __
WESTINGHOUSE PROPRIETARY CLASS 3 - . i. UNION ELECTRIC CALLAWAY PLANT NSSS UPRATING - 3579 MWt LICENSING REPORT REFERENCES
'1) R. H. McFetridge, R. T. Marchese, and R. H. Faas, A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant, WCAP-10263, January 1983. 1
- 2) ULNRC-1207. Callaway OTA Licensing Submittal, November 15, 1985. l
- 3) ULNRC-1247, Callaway BART LOCA for OFA Submittal, January 28, 1986. <
I l 4) ULNRC-1470, Callaway V5 Licensing Submittal, Narch 31, 1987.
- 5) SLNRC-85-06, MSLB Superheat Effects on Equipment Qualification, April 4, 1986.
l
)
l l l ! I 4 5970s:1dM31187 l l
9 c OBJECTIVES The Callaway Plant is currently licensed to operate at a core thermal power of ~ 3411'MWt (3425 MWt NSSS). This report supports the Union Electric application-to the Nuclear Regulatory Commission for approval to operate the Callaway Flant at 3565 MWt core thermal power (3579 MWt NSSS). A safety evaluation of NSSS design, operations and analyses has been performed to provide the following information relevant to that application:
- 1. A description of the proposed change in the licensed power rating of the Callaway Plant.
I
- 2. An assessment of the impact of that change on NSSS equipment designs,
- safety analyses, and systems operations.
- 3. A technical basis for establishing that the proposed increase in power rating does not involve an unreviewed safety question in accordance with requirements of 10 CFR 50.59.
9 i l l l 5970s:1d/031a87 i
.-w .-
- - - - - - - - - - _ _ - . _ _ ______a
- l
SUMMARY
The proposed increase in the licensed power rating of the Callaway Plant has been reviewed in detail with respect to its impact on the following aspects of NSSS design and operation:
- 1. The consequences of accidents postulated in the FSAR.
- 2. The capability of systems and equipment to meet design bases specified in the FSAR.
.- 3. The capability of equipment to r.aintain structural integrity under f conditions defined in the FSAR. ,
L ' I 4. Definition of NSSS/ BOP safety-related interfaces. I
- 5. Operating limits and conditions contained in Technical Specifications that are impacted by the power rating increase.
L This review has demonstrated that the Callaway Plant is capable, in its
; present design configuration, of operating at the proposed power rating and ,
i- remains in compliance with the design criteria and safety limits specified in I
~
h the FSAR for NSSS systeins and equipment, provided the plant is operated in accordance with the Technical Specification changes proposed in Attachment 2 I of the Callaway Uprating Submittal. The review has verified the following:
- 1. The probability of an accident previously evaluated in the FSAR will not be increased.
- 2. The consequences of an accident previously evaluated in the FSAR will not be increased. .
- 3. The possibility of an accident which is different than any already evaluated in the FSAR will not be created.
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- 4. The probability of a malfunction of NSSS equipme-t important to safety, previously evaluated in the FSAR, will net be increased.
, 5. The consequences of a malfunction of NSSS equipment important to safety, 'previously evaluated in the FSAR, will net be increased.
- 6. The possibility of a malfunction of NSSS 'equipme-t important' to safety, different froc any already evaluated in the FSAR, is not created by operation at the uprated power.
- 7. 'The margin of safety as defined in the bases to any technical specification will'not be reduced by operation at the uprated power.
~
Therefore, it has been concluded that operation of the Callaway Plant at the increased power rating does not reduce the NSSS safety sargins, and does not involve an unreviewed question.as defined by 10 CFR 50.59.- m 5970s:1d/031887 iii l
\
a. SECTION 1 INTRODUC110N I Union Electric has conducted a program to increase the electrical output of l'
- the Ce11away Plant. The current phase of the program is directed toward
. gaining approval from the USNRC to operate the plant at a slightly increased power level. At present, the Callaway Plant is licensed to operate at a core thermal- power rating of 3411 MWt. Union Electric is applying for an amendment to' the operating license to permit operation of the Callaway Plant at 3565 MWt (core thermal power),. an increase of 4.5%.
J As a part of the program to uprate the Callaway Plant, Union Electric authorized Westinghouse to perform a review of the NSSS systems and equipment designs to verify their capability to meet requirements for operation at 3579 MWt. The _ review was conducted in accordance with groundrules and criteria put forth in the Westinghouse topical report WCAP-10263, A Review Plan for i Uprating the Licensed Power of a Pressurized Water Reactor Power Plant (Reference 3). (This WCAP methodology was followed by both North Anna and Salem for their recent core power upratings.) A sumary of the guidelines used in the NSSS design review follow: i .' . F 1. Scope of Review The review encompassed all aspects of the Callaway NSSS design and operation which were impacted by the power increase.
- 2. Safety Review Acceptance Criteria ,
- NSSS designs have been reviewed at the uprated power level to verify continued compliance with licensing criteria and standards currently required by the Callaway operating license. In addition, a review has I been made as defined in 10 CFR 50.59 to identify any potential enreviewed safety question that might occur as a result of the l increased power rating.
I 5970s.1d/031a67 1-1 , - _ _ - - _ _ = _ - _ _ - _ _ _ _ _ b
I-
- 3. : Structural Review Acceptance Criteria-The< structural design of NSSS equipment'was reviewed at the increased power rating to assure that compliance has been maintained with industry codes and standa'rds that applied when the equipment was' originally built.
- 4. Functional Capability' A review has been'made.to verify that NSSS components and systems will-continue to meet the functional requirements specified in the FSAR at the increased power rating.
1
- 5. . Analytical Techniques r
Current NRC approved analytical techniques have been used for analyses { performed at the increased power rating. ,i
- 6. Balance of Plant Interfaces Information provided by Westinghouse to other design groups has been reviewed and revised when impacted by the increase in power rating.
]- Although Union Electric is applying for a license amendment to operate the dallaway Plant at 3555 MWt (core thermal power), many of the Callaway FSAR analyses and evaluations have already been performed at the engineered safety features design rating of 3579 MWt.. To snaintain a consistent basis between information reported here and that reported by reference to the FSAR, this evaluation of NSSS capability has also been performed at 3579 MWt. When the term *uprated power" is used in this report, it should be understood to mean 3579 MWt.
i i m e..teros2ss7 1-2
e o .. e l SECTION 2 l COMPARISON OF PARAMETERS-At-the present time, the Callaway Plant is licensed to operate at a core - L thermal power rating of 3411 MWt (3425 MWt NSSS). This amendment application requests approval to operate the plant at a power' rating of 3565 MWt (3579 MWt-NSSS). The calculated secondary side steam pressure at these conditions is 950 psia. Table 2-1 contains a summary of Reactor Coolant System design parameters for the originally licensed conditions of the Callaway Plant, as well as parameters calculated for the increased power rating. A comparison of the two sets of parameters follows:
- 1. NSSS Power The requested changes would raise the NSSS power from its current t level of 3425 MWt to 3579 MWt, an increase of 4.5%. This relates to an increase of the licensed core power from 3411 MWt to 3565 MWt.
6
- 2. Reactor Coolant Flow and Tube Plugging At uprated power conditions, the Callaway Plant will operate at i 3579 MWt with a range of 1000 psia to 950 psia steam pressure with 0%
f' to 10% of the steam generator tubes plugged. Table 2-1 lists primary plant parameters for the uprated power operating conditions, for both 10% and 15% Steam Generator Tube Plugging. The 15% steam generator tube plugging parameters constitute the conservative analytical basis for all of the Chapter 15 FSAR uprating analyses and evaluations except for the events associated with an increase in heat removal by the secondary system (Chapter 15.1) and those associated with an increase in RCS inventory (Chapter 15.5). The analytical basis for these events was 10% steam generator tube plugging. These events were analyzed at the uprated power level for Cycle 2 0FA licensing report ss70. w onss7 2-1
. o and explicit reanalyses were not required for the VANTAGE 5 licensing i j
report. The ~uprating licensing amendment assumes a 10% tube plugging . level.
- 3. Reactor Coolant Temperatures
, i Reactor coolant temperatures for the 3579.NWt NSSS power rating de not differ:significantly from those for the current 3425 MWt NSSS power
. level.. However, the higher power level and redu:ed flow rate are reflected by a slightly greater temperature rise in ine coolant as it passes through the reactor vessel (see Figure 2-1). ;
I
- 4. Steam Pressure and Temperature I
Operation at 3579 MWt NSSS requires an increase in stcam generator ! heat transfer rate, which is'obtained by increasing the temperature difference between r'eactor coolant and secondary plant steam and by . i increasing the steam flow rate. Since reactor coolant temperatures l for 3579 MWt NSSS operation are nearly the same as those for 3425 MWt NSSS operation, it follows that the higher power rating will be obtained at a lower steam temperature. Steam pressure (saturation
- pressure at the steam-temperature) will be correspondingly lower.
. j
- 5. Steam Flow- !
)
Steam flow at the 3579 MWt NSSS conditions has increased over the 3425 NWt NSSS conditions roughly in proportion to the thermal power increase. Comparison of the parameters given in Table 2-1 indicates that the operating i conditions proposed for future 3579 MWt NSSS operation do not vary significantly from those at which the Callaway Plant is currently operating. c. Figure 2-1 is a graphical comparison of the reactor vessel cold leg, hot leg, and vessel average temperatures as a function of power for both the current ss70..into31ss7 2-2
O , TABLE 2-1 CALLAWAY POWER CAPABILITY PARAMETERS Originally Licensed Uprated Uprated Power Power Power ' (10% SGTP) (15% SGTP) NSSS Power, NWt 3425 3579 3579 Reactor Power, NWt 3411 3565 3565
~he mal Design Flow, Loop gpm 95,700 93,600 93,600 Reactor Flow, Total, 106 lbm/hr 142.1 139.4* 139.4*
Reactor Coolant Pressure, psia 2250 2250 2250 l
- l h Reactor Coolant Temperatures, 'F ,
Core Outiet 621.4 623.7 623.7
.Y' Vessel Outlet 618.2 620.0 620.0 Cere Average 591.8 592.2 592.2 Vessel Average 588.5 588.4 588.4 ; Vessel / Core Inlet 558.8 556.8 556.8 ; !. Steam Generator Outlet 558.6 556.6 556.6 Steam Generator Steam Temperature, 'F 544.6 538.4 537.0 .
r' Steam Pressure, p!Ha 1000 950 939 6 15.1 15.9 15.9 Steam Flow,10 lbm/hr Total I Feedwater Temperature, "F 440 446 446 2ero Lead Temperature, 'F 557 557 557 Percent Tube Plugging 0 10 15 Core Bypass Percent 5.8 6.3** 6.3** 9 ased on Tjg = 556.8'T and 2250 psia
" Increased bypass flow is due to VANTAGE 5 fuel design, not uprating.
ssro..sofoszsa7 2-3 j- - - - L-________-__________________.
*
- WESTINGHOUSE PROPRIETARY CLAS$ 3 620" / '
W HOT LEG TDPGATURE /
/
' /
610" y q
/ 4
/ i
/
600" /
/
/
/
- h. 590 " / \
O / '
/ ~ > l
- U3 / /
} O' Sgg -
/
/ i I
/
e / hJ 570 "
/ W AVDWE1DFSATURE i
- 0. '
i 2 i,LI
- . - 560 "
t r W COLD LEG TDFUtATLRE I, 550 - 1
" UPRATED ]
540 - - CURRENT i 530- : 20 40 60 80 100 l Figure 2-1. Reactor Coolant Tempe atures vs. Percent Rated Load ! I a 2*4 6970s 1d/031187
M25 MWt NSSS operating conditions and the proposed 3579 WWt NSSS operating
- enditions. 'This figure.shows that the Reactor Coolant System temperatures
'or 3579 NWt.NSSS operation do not differ significantly from those for the eriginal 3425 MWt,WSSS operation t!roughout the power range.
4 i e 0 g ss70.wos2ss7 2-5
y .. . l' i SECTION 3 l ACCIDENT ANALYSIS REVIEW 3.1 NON-LOCA EVENTS All FSAR Chapter 15 Non-LOCA safety analyses applicable to the Callaway Plant and performed by Westinghouse have been analyzed or evaluated in support of the uprated NSSS power of 3579 MWt. Results of these analyses and evaluations are presented in the Callaway 0FA (Ref. 2) and VANTAGE 5 (Ref. 4) fuel licensing submittal reports. All applicable safety anelysis acceptance criteria have been satisfied. I
.. 3.2 LOCA EVENTS ~
Both the Large and Small Break LOCA ECCS analyses applicable to the Callaway 1
- Plant and performed by Westinghouse have been analyzed at the uprated NS$5 j i
thermal power of 3579 MWt. Results of these analyses are presented in the- ) Callaway 0FA (Ref. 2 and' 3) and VANTAGE 5 (Ref. 4) fuel licensing submittel reports. All applicable safety analysis acceptance criteria have been I j satisfied. 3.3 HOT LEG SWITCHOVER . f An analysis has been performed for the Callaway Plar.t to determine the maximum 1 boron concentration in the reactor vessel following a hypothetical LOCA. This analysis considered Callaway with an uprated NSSS thermal power rating of 3579 l NWt and maximum boric acid concentration of 2100 ppm in the RWST and ] accumulators and 2000 ppm in the RCS. The analysis considers the increase in boric acid concentration in the reactor vessel during the long term cooling phase of a LOCA, assuming a conservatively l small effective vessel volume. This volume includes only the free volumes of l the reactor core and upper plenum below the bottom of the hot leg nozzles. This assumption conservatively neglects the mixing of boric acid solution with directly connected volumes, such as the reactor vessel lower plenum. The calculation of boric acid concentration in the reactor vessel considers a cold q ss7o. w ostas7 3-1 (
h .- , L leg break of'the reactor coolant system in which steam is generated in the
. core from decay heat while the .coron associated with the borie acid solution is completely separated from the steam and remains in the effective vessel l volume.
l The results of the analysis show that the maximum allowable boric acid concentration of 23.53 weight percent established 'by the NRC, which is the boric acid solubility limit less 4 weight percent, will not be exceeded in the vessel if het leg recirculation is initiated 18 hours af ter the LOCA inception. The operator should reference this switchover time against the reactor trip /SI actuation signal. The typical time interval between the accident inception and the reactor trip /S! actuation signal is negligible when compared to the switchover time. Frocedures philosophy assumes that it would be very difficult for the operator j to differentiate between break sizes and locations. Therefore one het leg switchover time is used to cover the complete break spectrum. 3.4 HYDRAULIC FORCES The effect of the uprating on the LOCA Hydraulic Forcing Functions was l evaluated. The uprating effect was shown to result in an approximate 2% increase in the peak X-directional force actin'g on the vessel, with a - 6 j resulting force of approximately 6.952 x 10 Lb .f The LOCA Hydraulic Forcing Functions are based on a break area of 144 square inches. However per WCAP-960, Callaway's actual maximum break area is 73 square inches. The
)
application of a more conservative break area of 80 square inches to the analysis results in a decrease of approximately 25% which more than offsets I the effect of the uprating. Therefore, the effects of the uprating in combination with the credit for the break area reduction result in Hydraulic Forcing functions which are bounded by the original analyses. i i l 5970s;1d/031887 3-2 1
l I
.I SECTION 4 NSSS COMPONENTS INPACT -l i
4.1 BASIS FOR' EVALUATION The mechanical design of NSSS equipment has been' reviewed to assure that 1 j structural integrity of the plant will be maintained under conditiens i specified in the FSAR when the Callaway Plant is operated at the uprated i power. The review was perforced in accordance with the following guidelines: l f
- 1. The review encompassed sil aspects of the Callaway Plant NSSS {
equipment mechanical design which were impacted by the power increase, k
- 2. The review was per'ormed in accordance with licensing criteria and .
standards currently applicable to the Callaway Plant, z
- 3. Equipment sechanical designs were evaluated against the original industry design codes and standards to which the equipment was built, Current techniques- have been used for those analyses required in the l 4.
i course of the NSSS equipment review. . lt- . , These are the criteria put forth in WCAP-10263 as a basis for reviewing the I mechanical design of NSSS equipment during an uprating safety evaluation. ( 4.2 EQUIPMENT REVIEWS 4.2.1 Reactor Vessel To assess the impact of the uprating on the reactor vessel design and operation, the vessel stress report and fracture mechanics analyses were reviewed. This review veri'ied that the existing reactor vessel stress analysis bounds the uprated conditions, so the reactor vessel r6 mains in comp 1'ance with the currently applicable codes and standards. l 5970s:1d/c31887 4-1
A separate review was performed to assess the reactor vessel fracture mechanics. It was based on-the following inputs: l
- 1. The end-of-life fluence levels were calculated for the uprated conditions and low leakage fuel. Assuming 32 effective full power years, the total fluence _at the end-of-life at the reactor vessel 19 2 inner radius is 1.95 x 10 n/cm ,
- 2. The original design transients applicable for the Callaway Plant are unchanged.
- 3. Generic fracture mechanics evaluations have been performed at a power level that bounds the uprated power to evaluate the effect of a large I steamline break, large LOCA, and small LOCA transients on reactor T vessel integrity. Theseevaluationsindicatethattheproposedpowek, increase will not significantly change the results of the reactor' ves'sel integrity evaluation.
" Review of the reactor vessel fracture mechanics has indicated that operation at the uprated power condition is bounded by the original-design criteria.
4.2.2 Reactor Internals The reactor vessel inthnals review for the uprated conditions included three separate areas: a therr.a1 assessment, a hydraulic assessment, and a j f
' structural assessment. The review indicated that the original reactor internals design remains in compliance with the current FSAR design l requirements when operating at the uprated power conditions.
Since the design transients for the uprated power are bounded by those used for the original reactor internals design, only components which are directly influenced by the core radiation heat generation need be structurally j evaluated. These components are the baffle-barrel-former region, the upper { core plate and the lower core plate. Results of the reactor internals ' evaluations and analyses are summarized as follows: l s97oe 16/031st7 4"2 I 1
I p
- 1. The increase in the core bypass flow is not associated with the power increase, but results from the mechanical design of the VANTAGE 5 fuel.
l l
- 2. Because of the decrease in fluid density at the uprated conditions, l hydraulic lif t f'orcee and resultant loadings are bounded by those at f the conditions used in the original reactor internals design. !
- 3. The potential for flow induced vibrations is not increased.
- 4. Stresses and fatigue usage factors for components in the baffle
- barrel-former region of the internals are bounded by the original analysis.
- 5. Stresses and fatigue usage factors for the upper and lower core plates E are bounded by analyses performed for other plants using the same core plate design.
Sased on these evaluations and analyses, the Callaway reactor internals remain in compliance w ti h the requirements of all applicable design criteria as defined in the FSAR.
- I, 4.2.3 Reactor Coolant Pumps
. [. -
)
Review of the reactor coolant pump design indicated that operating. conditions 7 j for 3579 HWt operation are bounded by the original thermal and structural design analyses.
)
i 4.2.4 kntrolRodDriveNechanisms Review of the control rod drive mechanism design showed that operating conditions for 3579 MWt operation are bounded by the original thermal and structural design analyses.
\
ss7o.wonsa7 4-3
)
1 ___ _ ________ _ _ _ _ l
4.2.5 Reactor Coolant Piping l The stress analyses for the reactor coolant loop, reacto* coolant loop bypass, and pressurizer surge line piping were reviewed for the 3579 MWt operating
- enditions. The review demonstrated that piping stresses and support loads at the uprated conditions remain in compliance with the requirements of all applicable design criteria as defined in the FSAR'.
4.2.6 Pressurizer P.eview of the pressurizer design verified that operating conditions for 3579 NWt NSSS operation are bounded by the original thermal and structural design analyses. As indicated in Section 5.2.1, review of the Reactor Coolant System design has established that the existing pressurizer safety valves and power operated relief vehes are adequate for operation at the uprated conditions. l 4.2.7 Steam Generators Modifications to the model F series steam generator stress report have been
,' made using a set of operating parameters which bound conditions for the -
Callaway Plant at 3579 MWt NSSS operation. The ASME Boiler and Pressure j [ i Vessel Code, Section III 1971 Edition with Addenda to and including Sumer r
- 1973, was used te :fetermine acceptable states of stress for the components.
The evaluation established that the model F series stear generator stress report satisfies all applicable ASME Code requirements when updated to the enveloping set of plant operating parameters. An appendix to that report shows that the Callaway operating parameters for 3579 MWt NSSS are conservatively bcunded by the enveloping parameters. 4.2.8 Reactor Coolan,t System Supports Uprating of the Callaway Plant to 3579 MWt has a negligible impact on the Westinghouse supplied portion of the Reactor Coolant System supports. It was l
)
ii ss70. 1ero2:se7 4-4 i
b-_ . . . f established in Section 4.2.5 that design loads on the supports remain in complince with the requirements' ef all applicable design criteria as defined in the FSAR. 4.2.9 Nuclear Fuel Fuel performance evaluations completed for each fuel region demonstrate that the design criteria will be satisfied for all fuel in the core under the plenned operating r.enditioas. These fuel rod design evaluations were performed using the currently NRC-approved model PAD 3.3 (WCAP-8720, October, 1976). 4.2.10 Auxiliary Systems Components
~he Callaway auxiliary systems components supplied by Westinghouse were evaluated based on a comparison of the original design requirements to the systems design requirements at the uprated conditions. In each case the ~
conditions used in the original design enveloped those required for operation of the Callaway Plant at 3579 MWt NSSS.
4.3 CONCLUSION
S l
? Review of Callaway equipment designs that are impacted by the power uprating ,
b has shown that in most cases requirements for operation at the higher power
- are enveloped by either the original Callaway design, or by the generic component design. In a few cases, it has been necessary to perform additional design calculations to verify the capability of a component for operation and compliance with the original design codes and standards at the uprated conditions. In every case, however, it has been shown that the NSSS equipment originally provided for the Callaway Plant is capable of operation at 3579 MWt NSSS without modification.
f 5970s id/032sB7 4-5
SECTION 5 NSSS SYSTEMS REVIEW 5.1 BASIS OF EVALUATION NSSS systems designs have been reviewed to verify that they will remain in compliance with the functional requirements specified in the FSAR when the Callaway Plart is operated at the increased power rating. That review was performed in accordance with the following guidelines:
- 1. The review encompassed all aspects of the Callaway NSSS design and operation which were impacted by the power increase.
- 2. The review was performed in accordance with licensing criteria and standards which currently apply to the Callaway Plant.
- 3. Current techniques heve been used for those analyses required in the course of the NSSS review.
Theso are the criteria put forth in WCAP-10263 that apply to the review of the design and operation of NSSS systems and components. 5.2 SYSTEMS EVALUATION ( I 5.2.1 Fluid Systems l 1 Westinghouse has evaluated the impact of uprating the Callaway Plant to 1 j 3579 MWt on the reactor coolant system and the auxiliary fluid systems listed in Table 5-1. The only system which is impacted is the performance of the residual heat removal system for plant cooldown. As a result of the higher j decay heat associated with the power uprating, plant cooldown times will increase slightly. Operation of the Residual Heat Removal System is initiated at a reactor coolant temperature of 350'F, approximately four hours af ter ) reactor shutdown. The original 16 hour plant design cooldown time from 350'F l 5970s.1d/0318s7 5-1 i L !
TABLE 5-1. SYSTEMS REVIEWED FOR UPRATING IMPLEMECATION
- 1. Reactor Coolant - (BB)
- 2. Chemical and Volume Control - (BG;
- 3. Residual Heat Removal - (EJ)
- 4. High Pressure Coolant Injection - (EM)
- 5. Accumulator Safety Injection - (E:)
- 6. Containment Hydrogen Control - (ES) i 6
l , [ l i 1 i j l l l l 597De 16/031687 5-2
, 7 . g .i
- g ,\~
1 l' : o- , , A
~
However, this increase in plant to 140'T will new reqMre 19.3 hours.
} '
cooldown time is'not considered to be significant; safety requirements are still satisfied with respect to a single train cooldown under ac(ident conditions'. NochangeshavebeenrecommendedfortheRRRheate$changersor The changes in plant for the required tubeside or she11 side flow rates. cooldown performance for'the residual heat removal system as a result of the ! 1 uprating have been reflected in Table 5-2. [r j Changes in pressure, temperature, and flow rate around the reactor coolant h, loopsaresosmallthatthere'isessentiallynoimpactonpressurizerspraf
' capability, pressurizer safety and relief valve' discharges, pressurizer surge line' capability,andRTDbypass!delaytime.
i5.2.2 Cn0!rolSystems 4 ,
'l
,t 2
T s b t B'ased on Reatter Coolant System opyrating parameters for the uprated power conditions, stuMen were performed to assess operating margin and control i . t, system capability. The capability of th4 WSSS control systems (e.g., rod
' control, steam damp, pressurizer pressur's, and level control) was found to be Minor changes to the Tgf adequate for operation +t the uprated power.
- i program which feeds the rod control, steam dimp control, and pressurizer le cc,ntrol have been identified for operation at.the uprated power based onT these
[b studies.
- s
't
- 5.2.3 Protection Systems 9 x
k The protection system setpoint changee necessary for the uprating (OTAT and s The VANTAGE 5 3 DPAT) were incorporated as part obCycle 2 (Ref. 2). licensing submittal (Ref. 4) contrins additional chaNMs to Technical , Specification Table 2.2-1 which are due to the combined effects of VANTAGE 5
'M fwl, the uprating, and 15% steam generator tube plugging, s
, s 5t.2.4 AMSAC ,
The AMSAC oherating bypass setpoint, C-20, is currently sat at 40% of nomina
- turbine load or 1370 MWt. This setpoint is based on analyses which 5-3 s970s:1d/031as7 m
., a:
l TABLE 5-2 CALLAWAY PLANT C00LDOWN FOR 3579 Wt Reactor RHR Heat Exchanger Parameters Time After Shutdown Coolant , Heat Lead Per Heat Exchanger on Component Cooling Water Temp 'F 106 Btu /hr Inlet Temp. (*F) Outlet Temp. ('F)- (Hours) o. 4.0 350.0 300.0 118.0 116.6- 147.6 5.0 115.0 116.0 146.3 6.0' 250.0
.205.0 93.5 112.4 137.0 7.0 183.0 '74.5 109.2 128.8 8.0 172.8 65.7- 107.7 125.0 9.0-167.5 61.3 106.8 123.0 10.0 164.5 -58.7 106.5 121.8 11.0 162.4 57.0 105.9 120.9
.12.0 160.8 55.7 105.6 120.3
, 13.0 155.9 51.6 104.9 118.4 14.0 150.0 46.5 103.9 116.2 15.0 147.0 44.0 103.4 115.0 16.0 145.3 42.6 . 103.1 114.3 17.0 '
144.1 41.6 102.8 113.8 18.0 143.1 40.9 102.6 113.4 19.0 142.3 40.3 102.4 113.0 20.0 141.5 39.6 102.3 112.7 21.0 140.B 39.0 102.2 112.5 22.0 140.1 38.4 102.1 312.2 23.0 140.0 38.2 102.1 112.2 23.3 5970s:1df01%67 5-4 i
- -__-____-._____._____________,______m_ _ _ _ _ _ _ _
.. c demonstrate that in the event of a loss of heat sink ATWS occurring at that power level, the resulting peak reactor coolant system pressure would be below the Level C acceptance criteria of 3200 psig and that bulk RCS boiling would not occur. The analyses were performed for a limiting plant configuration as described in WCAP-8330 and in Westinghouse letter NS-TNA-2182 "ATWS Submittal" from T. M. Anderson to S. Hanauer (NRC), Dec., 1979. The Callaway Plant with a power. level of 3425 MWt is enveloped by this limiting plant configuration.
At the' uprated power level conditions of 3579 MWt, Union Electric will set the C-20 setpoint at (or below) the same hbsolute power level,1370 MWt, or approximately 38.3% of the uprated turbine load.
5.3 CONCLUSION
S Review of NSSS systems design and operating capability has verified that they will remain in compliance with the functional requirements specified in the FSAR when the Callaway Plant is operated at the increased power rating. A few system parameters and setpoints have been revised to reflect operation at the higher power, but those. operating conditions have been shown to be within the design capability of the systems as they currently exist. i
\
se7o..wonaa7 5-5 l L_____--____-
p. ( ,
- -y WESTINGHOUSE PROPRIETARY CLASS 3
\ g E , SECTION 6 NSSS/ SOP lNTERFACES. 6;l -1 TRODUCT10N N
, To coordinate the NSSS review with the Balance of Plant (BOP) review, a program was established to examine plant design data in those areas where the This section presents power uprating could have an impact'on the BOP design.
In addition, Westinghouse , investigated the l the results of that evaluation. L impact on the performance requirements for the systems identified in t Table 6-1. The review indicated that the current system performance ! requirements remain applicable at the uprated conditions. i - 6.2 NASS AND ENERGY RELEASE DATA f The originel Loss of Coolant Accident data for containment integrity evaluations was based on an NSSS power rating of 3579 MWt. This data remains j applicable since this is the uprated power. The original steamline break mass and energy release analyses for containment integrity were performed for various power levels up to 3425 WWt NSSS power. Additional cases had to be performed at 3579 MWt NSSS power. These calculations were performed using f j
- current Westinghouse methodology. As such, superheated steam is modeled and '
no' credit is taken for entrainment. All other modeling assumptions are consistent with the current FSAR Section 6.2.1.4 text. Results are discussed in Appendix B of Attachment 5 to the Callaway Uprating submittal. 4 A separate review has been performed to assess the effects on equipment qualification of steamline breaks outside containment with superheated , l blowdowns. This review used the mass and energy release information developed l by Westinghouse and reported in W AP-10961-P. The effects of uprating to 3579 MWt were considered in this EQ review. (Ref. 5.) 6.3 AUXILIARY FEEDWATER SYSTEM ; The design basis transients end accidents that govern the Auxiliary feedwater System (AFS) performance requirements were specifically performed as part of ss70.wom:7 6-1
.; s-
}- I TABLE 6-1 3 SYSTEM PERFORMANCE REQUIREMENTS REVIEWED FOR UPRAT*NG IMPLEM e i 1 .
- 1. Reactor Wakeup Water - (BL) 1.
Berated Refueling Water Storage - (BN) l ! 2. 1,
- 3. Containment Spray - (EN)
- 4. Containment Atmosphere Control - (GR)
- 5. Containment Purge - (GT) 3
- 6. Gaseous Radwaste - (HA)
- 7. LiquidRadwaste-(HB)
- 8. Solid Radwaste - (HC)
- 9. Boron Recycle - (HE)
Reactor Protection System - (SB) i' 10. i
- 11. Reactor Instrumentation - (SC) 4.
l
- 12. Ex-CoreNeutronInstrumentation-(SE)
- 13. Reactor Control - (SF)
- 14. IncoreNeutronInstrumentation-(SR) {
1 1 1 1 l l l 6-2 { ss7o. w ostas7 l
~~^~-"---_--._.,_m. , _ ,
j j I t the Callaway Cycle 2 reload' analysis and it was demonstrated that the existing l AFS requirements are unchanged by the uprating. l 6.4 RADI ATION SOURCE TERP.S
' The radiological source terms were evaluated as part of the Cycle 3 reload analysis at the uprated power level. (Ref. 4) 1 6.5 CCHPONENT COOLING WATER INTERFACE REQUIREMENTS Interf ace requirements for the component cooling water system have been ev'aluated. The only heat load which changes for Westinghouse supplied equipment is the load to the residual heat removal heat exchanger. Operation of the residual heat removal system is initiated at a reactor coolant temperature of 350*F approximately four hours after reactor shutdown. At this j peint in plant cooldown, both RHR heat exchangers are capable of removing a heat load composed of decay heat generated by the reactor core plus heat input from the reactor coolant pumps, plus sensible heat released by a 50'F/hr plant cooldown. The increased power level from the uprating means that the core decay heat load wii' increase in proportion to the core power rating.
For " Plant Shutdown at Four Hours," the heat 1.oad for each of the residual ' 6 heat exchangers should increase from the current value of 116 x 10 BTU /HR , 6 per heat exchanger to 118 x 10 BTU /HR per heat exchanger. Note that the fractional increase in the residual heat removal heat load is less than the fractional increase in core power because the core power affects only a pertion of the residual heat removal heat load. The component cooling water requirements for the RHR heat exchangers remain j unchanged. This is censistent with calculated plant cooldown performance for the uprated con'itions under the assumption that she11 side cooling water flow rates would remain unchanged.
}
There are no other impacts c component cooling water from Westinghouse designed systems. { 4 ss W :1d/o31187 6-3 1 a
.. 9 ' J.- l
. Attachment 5, Appendit B- --
Section I
.'.d. :
ULNRC-1471~ L March 31,~1987 a I i CALLAWAY UPRATING SUBMITTAL ATTACHMENT 5: APPENDICES APPENDIX B BOP UPRATING.- LICENSING REPORT , s SECTION I - BECHTEL REVIEW s 8 W h 1
q .- BOP UPRATING LICENSING REPORT I. INTRODUCTION The Callaway plant is currently licensed to operate at.a core power level of 3411 MWt. Union Electric has proposed uprating the plant to the 3565 MWt core power level. (3579 MWt NSSS) . In order to determine the impact of uprating on the existing design and identify the required modifications, if any, a feasibility study was conducted. The feasibility study demonstrated that no hardware modifications are required to achieve the uprated NSSS power level of 3579 MWt. For the feasibility study Bechtel used the valves Wide Open (VWO) heat balance (Figure 1) as the basis for reviewing system and. component performance. For -final implementation, Bechtel evaluated performance using the VWO heat balance, but with~the modified pressure, temperature, and flow from cases shown in Table 1. These cases reflect extreme conditions which Liight be encountered should.the full 3579 MWt be obtainable (as opposed to being limited to the VWO heat balance power of 3562 MWt). Table 2 contains a breakdown of systems that Bechtel reviewed for final implementation. System review includes confirmation that all design criteria envelop the uprating parameters; confirmation that all stress analyses are adequate for the new conditions; and that the components meet new operational criteria. This report summarizes the results of the final implementation ,
!- review.
e I t 1 l s 1 J
-y .
s i .; II.
SUMMARY
AND CONCLUSION
'The BOP systems'1isted in Table.2 were reviewed to determine the' impact of.uprating the:Callaway Plant from the current licensed core power level.of 3411 MWt to 3565.MWt. The. review consisted of comparing the existing design with the performance requirements at the uprated power level and determining if modifications were-required to the plant and.the documentation.
> Bechtel evaluated performance using-the.vWO heat balance but with the modified. pressure, temperature, and flow from cases .. shown in Table 1. These cases reflect extreme conditions which might be encountered should the full 3579 MWt.be obtainable (as opposed to being limited to the VWO heat balance power of-3562-MWt). Operating data was also reviewed to assure that.the plant will be able to' achieve the uprated power level. This was:necessary-because the operating data could be significantly different from;the design' conditions. The impact on containment pressures and temperatures following k a main steam line break was evaluated:- four (4) main steam line break cases at 102% of uprated power level were analyzed i, and the effect on equipment qualification was checked. The turbine-generator system is designed.to operate at 3562 Mut power. The performance of the system will be monitored closely by U.E. between 3562 MWt and 3579 MWt power. In conclusion, all the secondary side systems were reviewed, and it was concluded that, with the exception of the turbine-generator system, they have the capability to' function properly at the uprated power level c'f 3579 MWt NSSS power without any
- modifications to the existing design. +-
e W M
s L
- f. Y 'III; SYSTEM' EVALUATION l .:
I ' Main Feedwater System L The' main-feedwater system consists.of theLfeedwater pumps, high pressure 1 f eedwater-heaters, feedwater flow control valves, and associated' piping. . The feedwater: pumps.take suction from.the l condensate: system and'from the heater drain pumps and discharge through the'high pressure heaters to the steam generators. At'an uprated. power level of 3579 MWt, the extreme: conditions for the main feedwater system would occur when no steam generator-tubes are plugged. The. temperature and6 flow under this condition would be 446'FThe (max.) and'15.96x10 lb/hr (17,743-gpm) respectively. design' temperature for.the 6 feedwater system is-444.5'F, and the design flow is 15.85x10 v Ib/hr (17,6 2 0 . gpm) . . Therefore, at an uprated power level,'the feedwater system,will see a rise in temperature of 1.5'r and a 9 flow increase.of less'than 1 percent from the previous VWO
- p. design flow..
I The effect of such'a small inceease in flow on the flow. I velocities, system pressure. drop and the high pressure heater performance will be negligible. For the small change in flow, i there will be.no impact on the performance of the pumps. These l-pumps, including their turbine. drivers, have sufficient capacity to produce the uprated flow. .
- l
, The main feedwater bypass control valves are used until 20%
(approximately) power level is reached and, therefore, will not i be affected by the uprating. At higher power levels, the
- f. feedwater flow is controgled by feedwater control valves which have a range of 0-4.8x10 lb/hr, well within the uprated flow of 3.99x10 lb/hr. .
Condensate Systems A) The condensate system delivers water from the main L condenser hotwell to the steam generator feedwater pumps. The system: consists of the main condenser, condensate pumps, low pressure feedwater heaters, and the associated piping and instrumentation. The condensate system is designed to handle 50 percent greater than normal flow. This was done to permit full feedwater flow , upon loss of heater drain pump flow. Therefore, a small increase-in flow of less than I percent due to uprating will not have any impact on this system. B) The condensate storage and transfer system supplies or receives condensate, as required, by the condenser hotwell level control system. At uprated level, there will be an , increase in the main ster.m flow to the condenser, but this.will j I result' l _3_
)
_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . . . _ _ _ _ _ ___ __ _ _ i
3 .. >
- in a proportional' increase in the condensate flow. Therefore, the hotwell level and the demand on CST system will not change, due to uprating.
C) In order to maintain the purity-of feedwater, the i condensaf:e goes through the condensate demineralized system
- prior to flowing to low pressure heaters and to the steam generator feedwater pumps. The condensate demineralized system (CDS) removes the corrosion products and condenser leakage impurities from the condensate.. At present, the CDS is designed for VWO steady state flow of'21,600 gpm. At.an uprated level of-3579 MWt NSSS power, thereThis will be an. increase will increase in main steam flow of less than 1 percent.
the flow to the condensate demineralized system by a similar small percentage and is judged to be acceptable. Auxiliary Feedwater System The auxiliary feedwater system provides a reliable source of safety grade water to the steam generators when the main feedwater system is not available. It may also be used-following the reactor shutdown to cool the reactor coolant system. As evaluated by Westinghouse, the design basis transients and accidents that govern auxiliary feedwater performance as well as its cafety grade water storage requirements for callaway Therefore, are , based on an uprated NSSS power level of 3579 MWt. uprating will not have any impact on the existing auxiliary feedwater system. Main Steam System The main steam system provides steam from the steam generators to the turbine generator system and other auxiliary systems for
'~
power generation. The' major components of this system are the main steam piping, power-operated relief valves, main steam safety valves, flow restricters, main steam isolation valves, main steam isolation bypass valves and the turbine bypass valves. 6 j The main steam system is designed for VWo flow of 15.85x10 ) lb/hr. At the uprated gower level lb/hr, which of 3579is MWt, an the system increase of less may have a flow of 15.96x10 than 1 percent. The increase in the velocities and the ) pressure drop will be insignificant due to this 16 creased flow. Therefore, the turbine throttle pressure will remain unchanged. p e f i f I i q l j
~
\ i Each main steam line is provided with five (5) spring loaded safet/ valves for overpressure protection and one power operatad relief valve. Higher plant rating would require higher relieving capecity. However, this will not be a problem. since the safety valves were sized based on the NSSS engineered safety features (ESP) design rating of 3579 MWt, and the increase-in steam flow is relatively small. There will be no change in the relief valve thrust forces. ; Fequired Capacit'y Actual Capacity at 3% Accumulation at 3% Accumulation Safety Valves No. (lb/hr) At 3579 MWt (1b/hr) 1 796,500 893,160 2 796,500 902,096 3 796,500 911,779
- 4 796,500 920,715 5 796,500 929,652
- y In order to limit the steam flow following a. steam line break, a flow restricter of 1.4 square feet at the steam generator outlet nozzle is installed. This restrictor will not affect uprated steam flow requirements.
In addition, the steam dump valves will have sufficient
, capacity to satisfy the uprating requirements, and the main
'. steam isolation valves, as well as the bypass isolation valves, will not be affected. .
1 . hE ' Feedwater Heater Extraction, Drains and Vents h The drains from the low pressure heaters No. 2, 3, and 4 4 [ cascade to the No. 1 heater, which drains to the main i condenser. Drains from the heater No. 7 cascade to the shell l of heater No. 6 which drains to the heater drain tank. The j heater drain tank also receives drains from the No. 5 heater and the moisture separator drain tank. Two 50 percent capacity ! pumps inject this drain flow into the suction of the steam generator feedwater pump. Level control valves on the drain lines of the low pressure heaters and on high pressure heaters No. 6 and 7 automatically maintain the normal wat6r level in the heaters. High pressure heater No. 5 drains by gravity l only. The increase in main steam flow of less than 1 percent due to plant uprating up to 3579 MWt power will result in correspondingly increased condensate flow of less than 1 percent. The change in the flow rates of the heater drains will be similar. As a result of the small increase in drain
- c. .
flow, the lerel control valves will modulate open wider to maintain the desired water level in the heaters. The set points of the level controllers will not be changed, and the drain system need not be modified due to uprating. . I The heater extraction system provides extraction steam from the high pressure turbine to high pressure heaters and from the low pressure turbines to the low pressure heaters. Also, the , scavenging steam from the moisture separator reheater (MSR) ) first and second stage reheaters is directed to feedwater heaters No. E and 7. The extraction flows will increase by less than 1-percent in proportion to increases in condensate / feedwater flows. The heater vent system removes the non-condensible gases from the shell side of the heaters. The plant uprating will not significantly increase the amount of these gases, and i therefore, this system will not be impacted.
) - Steam Generator Blowdown System l 1 The steam ger.erator blowdown system (SGBS), in conjunction wi'th ~
the condensate and feedwater chemical addition system, and the( condensate demineralized system, maintains the secondary side water chemistry within the NSSS specifications. At present, the extent of blowdown processing during full power ) operation is determined by the operator depending upon the type ! and level of operation. At uprated level, the processing capability cf the blowdown system will be utilized to the j extent required to keep the water chemistry witnin specification. In fact, the increase in feedwater/ main steam flow at uprated j(L l; level is so small (less than 1 percent increase) that no impact 'I on SGBS is expected. , Cooline Water Sy$tems 1
)
,o The component cooling water system (CCWS) provides cooling i water to several reactor auxiliary systems during a loss of coolant accident. The system serves as an intermediate barrier f between the intake cooling water system and the potentially j radioactive systems. The CCWS consists of four 100 percent ] capacity. circulating pumps, two heat exchangers, two surge j j tanks, one chemical addition tank, and a.ssociated piping, valves and instrumentation. . Increased heat loads from the spent fuel pool due to the uprating (and the use of VANTAGE 5 fuel) have been calculated. These loads have been evaluated and can be accommodated by the CCWS. i l I l i
a .
' Westinghouse's evaluation has shown that heat loads imposed on CCWS.from the residual heat renoval (RHR) heat exchangers are expected to increase slightly. The impact of this-will be that the time to cool the plant from 350*F to 340*F will now take 19.3 hours instead of 16 hours. No other impact on component cooling waterfaue to uprating, and no significant impact from the secondary side systems is expected.
The closed cooling water system (CICW) provides cooling to the various components inside the-turbine' building such as steam generator feedwater pump turbine lube oil coolers, condensate pump motor bearing oil coolers, heater drain pump motor bearing oil coolers, etc. Any increase in the heat load to the closed cooling water system will be small and.well within the system design limit. Note thgt the normal load on the closed cooling , water sgstem is 3.5x10 Btu /hr and that it is designed for 7.56x10 Btu /hr. The service water system provides a source of water to the non-essential auxiliary plant equipment at a maximum tempera-ture of 95'F. The essential service water system provides cooling water to the plant components which are required for the safe shutdown of the plant. The components served by these systems, except for the component cooling water heat exchangers, will not experience any significant increase in heat load. Since the CCWS increase was already considered in the original design, these systems will not be significantly affected by the uprating. The service and essential service water systems have sufficient margin to adequately supply any increase in flow required in the future due to conditions such as foulinc in the ! l heat exchangers. Containment Sprav System The objectives of the containment spray system are: 1) to reduce the containment atmosphere temperature and pressure in the event of a loss of coolant accident (LOCA), or a main steam line break (MSLB) inside the containment; and 2) to limit the offsite radiation levels in the event of a postulated LOCA. The pressure and temperature analysis for various main steam line breaks was done at an uprated level of 3579 MWt. The temperatures and pressures were found.to be within the maximum pressures and temperatures for which the containment perform-ance-was previously evaluated (Ref. Table 3). For LOCA, there is no increase in containment temperature, pressure, or radiation levels. Therefore, the containment spray system will not be affected by the uprating. l
's; t. t "T i y
I i l ( Peactor Makeup' Water System . k ) j The reactor makeup water' system-(RMWS) stores demineralized n water:and supplies it to various NSSS auxiliaries,.radwaste D systems, and the fuel pool cooling and cleanup system. An increased spent fuel. pool evaporation' rate'due to increased heat loads.has been calculated. The present RMWS' design provides sufficient makeup water for this demand. No-other l I change in'the demand for reactor makeup water is likely due to the uprating.-
- )
HVAC Systems The following systems were reviewed to determine the impact of
; uprating to'3579 MWt NSSS. power.
Central ~ Chilled Water System Au>:iliary Building HVAC Containment. Cooling System Containment Atmosphere Control System Containment Purge System There is negligible increase in heat-loads for the containment due to'uprating. The increase in heat loads due to any increase in the system operating temperatures is negligible. Radwaste Systems The gaseous radwaste, liquid radwaste,. solid radwaste, and the secondary liquid waste systems collect and' process waste prior to storage or discharge. This is a batch process, and if there is any increase in demand on these systems due to uprating, it ' will be slight and accommodated easily by.more-frequent process- - ing. The boron recycle system (BRS) receives the reactor coolant effluent for the purpose of processing and recycling. No change.is anticipated in the boron concentration or ' radioactivity level of the reactor coolant. Therefore, the BRS should not be affected by the uprating. Turbine Generator Review At present, the turbine generator is designed to operate at 3562 MWt and produce 3234 MWe (gross) . Union Electric considers it unlikely that this power level is limiting for the turbine generator and will carefully monitor the performance of the turbine generator system beyond 3562 MWt to the uprated power level of 3579 MWt. ____..-___._.____-___-__m____ .
- s" ;a-O' x~
L Main Generator System
- . i The main generator system. generates power.from a turbine- h
-generator which then is transmitted through an isolated p ase bus and'a main step-up transformer to'the offsite power system.
It also serves to step:down voltage through a unit-auxiliary
. transformer for normal operation of.the plant unit auxiliaries.
1,373,100 KVA at 0.90. power The' plant's generator is sizedfforTherefore, the plant's generator has adeq
-factor.
and margin for uprating the plant to 3579 MWt/1246 MWe.
'[
Excitation and Voltage Regulation The excitation and voltage regulation system provides the source of field current for excitation of the main generator and' control of the generator voltage by varying the field current to the exciter. The existing exciter.and voltage regulation system have- j, adequate capacity and margin for uprating the plant. U Startup-Transformer The startup transformer receives power from the offsite power i grid and steps down the voltage to supply the onsite electrical-
- The distribution system for startup and. shutdown of the plant.
l 5. startup transformer also serves as the source of power for one '
- load group.of the Class IE power system.
The startup transformer is fed from the offsite power gridThe transformer is s f(. - which is unaffected by the uprating. for full loading of the plant's equipment assuring adequate j capacity'to the plant auxiliaries. Lower' Medium Voltage System - 4.16 KV AC (IE)
.P The lower medium voltage system receives power from two 12/16 l
MVA ESF transformers and distributes it to the two redundant load groups in the Class IE system, The emergency loads to the system are not af fected by the plant uprating. Additionally, the ESF transformers are sized for full loading of the plant's equipment assuring adequate capacit y to the two redundant load groups. ' y Standby Generators The standby generators provide the power required for safe shutdown of the reactor in the event of loss of the preferred power source. l
The emergency loads to the standby generator are not affec ted by the plant uprating. All equipment loads required to safely shut down the plant are unchanged with the uprating. Load Shedding and Emergency Load Sequencing The load shedding and emergency load sequencing system provides for removing selected loads from the Class IE buses under degraded bus voltage conditions or 'upon generation of a SI signal and reloading the equipment in a predetermined sequence. - This ensures that the voltage and frequency of the' Class IE buses are not degraded due to heavy starting loads of actuated equipment. Load shedding and subsequent reloading of equipment onto IE buses is unaffected by the plant uprating. Miscellaneous Plant Power Systems The miscellaneous plant power systems include 13.8 KVAC (non IE), 4.16KVAC (IE and non IE), 480 VAC (IE and non IE), 250 VDC (non JE), 125 VDC (IE and non IE), instrument AC (IE and non IE), and uninterruptible AC. These systems serve the These balance-of-plant as well as turbine and NSSS systems. systems are required for normal operation and came shutdown, es applicable. The plant electrical systems listed above are sized for full load equipment operation, and, because the uprating of the plant does not require additional equipment, these systems are adequate for the uprating.
Process Sampling System 1 -
l The process sampling system provides representative samples of ( non-nuclear process fluids for analyses necessary for plant l operation, corrosion control, and monitoring of equipment and l system performance. J i l Plant uprating will slightly increase the temperatures of l certain fluids being analyzed (i,e., feedwater, steam generator blowdown, etc.). However, the system has more than sufficient { l capacity to absorb the increases in sample heat loads. Nuclear and Post Accident Sampling Systems i l The nuclear and post accident campling systems provide I representative samples of process fluids for radiological and j chemical analyses necessary for plant operation, corrosion control, and monitoring of equipment and system performance. Plant uprating will slightly increase the temperatures of certain fluids analyzed during normal and accident conditions; however, these systems have adequate capacity to absorb the projected increases in temperature and pressure.
L 6 IV. MAIN STEAM LINE BREAK (MSLB) PRESSURE / TEMPERATURE ANALYSIS FSAR Table 6.2.1-5611ists the 16 cases that were' analyzed' originally to determine the worst case containment pressures and temperatures'following a-main steam line break. .out of thesr cases, the following four cases were reanalyzed to
. determine the impact of plant uprating to 3579 MWt. l
- 1. Full double-ended rupture at'102 percent uprated power (Case 1). l
- 2. 0.60 ft 2 double-ended rupture at 102 percent uprated power (Case 2).
- 3. 0.80.ft 2 split rupture at 102 percent uprated power (Case 3).
Full double-ended rupture at 102 percent uprated power 4. (Case 16) assuming failure of'MSIV.
.The method and assumptions used The in.the 8' analysis percent are consistent revaporization of with Chapter 6 of the FSAR. The analysis condensate as allowed by NUREG 0588 was modeled.
was based on the new mass and energy release data provided by Westinghouse, p 2 split _ rupture was an iterative The analysis for the 0.8 ftFor-the uprated condition, the tir.e to reach Hi-1 set process. pressure of 6.0 psig and Hi-2 set pressure of 20 psig were This found to be 16.6 seconds and 61.6 seconds, respectively. information was provided to Westinghouse to3 obtain the total mass and energy release data for the 0.8 f t' split break case. a The reanalysis of the four cases described above indicated that - the maximum temperatures and pressures inside containment for the uprated-condition are similar to the ones calculated , earlier for the current power level of 3425 MWt. .t Figures 2 thru 9 provide the revised pressure and A summary of the temperature results is profiles for the uprated cases. given in Table 3. 1 f _ _ . __ _-_ _ _ _ _ _ _ _ _ _ - _ _ _
. 3-
~i V. EQUIPMENT QUALIFICATION From FSAR Table 6.2.1-2, the maximum pressure and temperature i inside'the containment at the' current power level of 3425 MWt
'l are 48.11psig and.384.9'F,'respectively. - Table 3 indicates that the te.mperatures and pressures;will increase at the upratedLpower-for7some of the main steam line breaks, but they' l
Lwill remain within the maximum levels specified above. Further, a' comparison was done between the temperature profiles of MSLB at the uprated. power level and the EQ envelope.given in Figure 3.11 (B)-3 of the FSAR. Only in one instance, for the-full double-ended rupture with no'MSTV (Case 16), the maximum x temperature of 352*F at 45 seconds exceedeo the EQ limit.. In order to confirm that equipment qualification will not.be-
~
affected, all equipment inside the containment was reviewed.
- The review indicated that the equipment is'either qualified for the EQ temperature and pressure envelopes (FSAR Figures-3.11 (B)-3 an'd 3.11 (B)-6) , as revised by the new case 16-parameters, or it was determined. that,, even during these peak '
conditions inside the containment (with consideration given to
- 1 the new case 16 results) , the actual conditions of the equipment will be within the limits for which it is qualif, led "Therefore, no (Reference 1).
will occur due to the plant uprating. x 4 t 1: e o
s
- VI. REVID? OF OPERATING DATA The secondary side operating data was takenThe when the plant feedwater and was operating at 99.66 percent of full power. For main steam data for each individual loop was evaluated.
some of the.feedwater heaters, minor deviations in temperatures
,and water levels were noted, but their overall performance was.
good. This was confirmed by the final feedwater Thesetemperature data indicate being very close to the expected values.no anticipated problems with ope uprated conditions. The evaluation of component cooling water and closed cooling water heat exchangers was performed based on the operating data I The closed cooling water heat exchangers will see provided.the bigger difference in operating heat load at uprated l conditions. The operating data indicated no significant l performance problem such as would be caused by extensive tubeit is our fouling. .Therefore,
~
perform adequately at uprated conditions. i 6 a
- - - - - - - - - - _ - ' ' ~ - - - - - - - - - - _ ~ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ - ' ' ' " - " . , _ _ _ __ _ _ _ _
- _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - .=. _ _ _ _ _
VII. EJTECT ON PIPE STRESS EVALUATION \ Thepipeskressanalysesofthesystemswereoriginallydone At 3579 using the VWO conditions at 3562 MWt NSSS power level. i MWt, the; temperature of. main feedwater will increase by only 1.5'F, and.the main steam pressure may decrease. For other systems also, the changes in pressures and temperatures will be negligible. The increase in flow in the systems, if any,i.will also be too small to impact pipe stresses of systems for which transient analysis was done. Considering the insignificant changes in system operating conditions and the conservatism in ! stress analyses and the pipe support design, no pipe stress reanalysis will be required due to uprating. m e i l {, f l
,0 VIII. HAZARDS ANALYSES Uprating the Callaway Plant from the current licensed core power. level of 3411 MWt.to a new licensed core power level of 3565 MWt can be accomplished without design changes in the BOP' systems._ Therefore, no changes to the facility will be made to achieve theluprated power level.- The following hazards analyses were reviewed in.the uprating study.
- 1) II/I Analysis: The existing stress analyses of the II/I piping systems will not be affected because no significant changes in their operating conditions will' occur. Also, no modification to the piping geometry of any kind will be.
made. Therefore, uprating'will not have any impact on the II/I evaluation.
- 2) Pipe Break and Jet Analysis: The changes in pressures and temperatures of safety-related main steam and main ! feedwater piping are'too insignificant to have any effect n on_the existing break locations. Current jet impingement analyses are unaffected as these analyses are generally conservative, and the increase in the flow rates are less p
than one '(1) percent. The operating conditions of cther F safety-related'high energy systems evaluated will'not change due to uprating. . Therefore, their pipe break locations and jet analyses will not be affected. For main steam line breaks inside containment, pressure and temperature analyses were done'and it was found that the maximum temperatures and pressures will remain within the peak pressure and temperature used in the evaluation , of containment performance. In addition, equipment inside [ I s- containment was reviewed, and it was confirmed that their environmental qualification will not be affected. .
}
- 3) Flooding: The only systems with an increase in flow rates j are main feedwater, condensate, and main steam systems. j This increase is less than one (1) percent and, therefore, will not have any significant impact on the flooding analysis in safety-related areas of the plant. 9 l
Moderate Energy Cracks: None of the moderate energy lines 4) will. experience any significant change in their operating conditions due to uprating. Therefore, the *No Crack Zones" as well as the evaluation done for the moderate ~ energy cracks will not be affected.
- 5) Missile: The increaseTherefore, in the speeds of fans will be uprating will not cause ,
insignificant if any. l any additional concern due to missiles.
- 6) Fire Hazards: The fire protection systems are independent of the plant power level and will not be affected by the uprating.
J l
.m sammun.manm e-
-----L.--,--,,,--_-_-_.-__. < - - - - - _ - - - - - - - - - - - - - , - - -
_ _ . , _ . _ - - _ _ - _ _ . = _ _ _ - _ _ . .. . . _ - _ __. ____
. e.
.a it' IX. REFERENCES "
,?(.
1, l' ) SLNRC L 86-02' dated 1/17/86, " Report of Independent Revies'of Environmental Qualification Programs to NUREG-0588",, , Volume 2 2) SLNRC B6-06 dat'ed 4/4/86, " Main Steam Line Break Luperheat' Effects.on Equipment Qualification" t, h' e
'f .
TABLE 1 CALLAWAY UPRATING PARAMETERS VWO Heat Uprating Uprating Uprating Parameter Balance (Case 1) (Case 2) (Case 3) 3562 3579 3579 3579 N555 Power (MWth) 0 10 15 S/G Tube Plugging (t) 0 Throttle Press. (psia) 975 975 925 914 Feedwater Temp. (F) 444.5 446 (1) 446 (1) 446 (1) Steam / Feed Flow (MLBM/HR) 15.85 15.96 (2) 15.92(2) 15.91(2) Generator Output (MWe) 1234 1246 (3) 1246 (3) 1246(3) l
,gTREMECASES
= 15.96 MLBM/HR Max. Steam / Feed Flow Max. Feedwater Temp. = 446 F Min. Throttle Pressere = 914 psia Max. Generator Output = 1246 MWe House Loads = 53 MW, 23 MVAR (4) l l Notes: (1) Assumed; based on extrapolating Feedwater Temp. vs. Power.
(2) Calculated as follows: I (Flow MLBM/HR) = (Power MWt) (3.412'4 BTU /W.HR) (h steam - h feto BTU /LBM) Where: hsteam=1191.4 BTU /LBM(Case 1
= 1193.1 BTU /LBM (Case 2
= 1193.5 BUT/LBM (Case 3 (assumes saturated steam at throttle pressure + 25 psi, 0.25%
moisture) h feed = 426.2 BTU /LBM (assumes 446 F, 1150 psia) (3) Maximum output; based on extrapolation to 3579 MWt and lower circulating water temperature. (4) Based on plant operating data; not expected to change significantly due to uprating.
- TABLE 2:
SYSTEMS REVIEWED FOR IMPLEMENTATION
~1) Main Steam .(AB) 2)- ' Main' Turbine (AC) 3)' Condensate - (AD)
.4 ) . Main Feedwater (AE)
'5) Feedwater Heater Extraction, Drains and Vents - (AF) _
- 6) Condensate Demineralized - (AK)
- 7) Auxiliary Feedwater - (.AL) -
8). Demineralized Water storage and Transfer - (AN)
- 9) Condensate Storage and Transfer - (AP)
' 10 ) _' Reactor Makeup Water - (BL)
- 11) Steam Generator Blowdown - (BM)
- 12) Borated Refueling Water Storage (BN) 13)' Steam Seals - (CA) -
- 14) Main. Turbine Lube Oil --(CB)
- 15) Generster H2 and CO2 - (CC)
- 16) Generator Seal Oil - (CD)
- 17) Stator Cooling Water - (CE)
- 18) Lube Oil, Storage and Transfer System - (CF)
.19) Condenser Air Removal - (CG)
- 20) Main Turbine Control Oil - (CH)
-21) Service Water - (EA) '
Y- 22) Closed Cooling Water -- (EB)
- 23) Essential Service Water - (EF)
- 24) Component Cooling Water - (EG)
- 25) Containment spray - (EN)
- 26) Auxiliary Turbines - (FC)
- 27) Central Chilled Water - (GB)
- 28) Auxiliary _ Building HVAC - (GL) .
29)' ' Containment Cooling - (GN) ,
- 30) Containment Atmosphere Control - (GR)
- j. 31) Containment Purge - (GT)
- 32) Gaseous Radwaste - (HA) i
TABLE 2< i-l- N ' c' ~ SYSTEMS REVIEWED FOR PHASE II '(IMPLEMENTATION)' . (cont'd) I r. L 33)' Liduid-Radwaste - (HB)
~
l'- 34). ' Solid Radwaste - (HC) . 35). Boron' Recycle - (HE)
> -36) Secondary Liquid Waste.- (HE);
- 37) Main Generator - (MA).
- 38) Excitation and Voltage Regulation - (MB)
.39) Startup Transformer - (MR)
'40) 4.16 KV AC (IE) -
(NB) f. Standby Generator - (NE)
' 41) i 42) 480 V AC (IE) ' (NG) -
- 43) 125 V DC (IE) -
(NK)
- 44) Instrument AC (IE) - (NN)
~
13.8 Kv AC - (PA) f 45)
- 46) .4.16 Kv AC - (PB)
- 47) 480 V AC - (PG)
- 48) '250 V DC -(PJ)
!* 49): 125 V DC'- (PK)
- 50) Instrument AC - (PN)
- 51) Uninterruptible AC - (PQ) l 52) Process Sampling - (RM)
I 53) Nuclear Sampling - (SJ) ,
.e g
~ A f I
.I TABLE 3 ..
SLW. ART OF RT.5M,Ts Tsar 9 t 31-2 31-3 dryout W Faas 9 t 31-1 (sec) (se,c) (psis f see) (#T $ see) (sec) (sec) case , 348. t 110. 2.13 12.86 78.5 186. full DE3 . 39.35 0 1800. 29.05 110. 164.
"-i 38'. 4 1800 318. 9 145. 2.69
, rsAR 358. t 310. 10.7 120.1 273. 355.6 2.6 it.2 DEB 36.79 9 1800. 113.7 280. 388.
a 37. t 1800. 372. t 310. 11.39 rtat 2.13 8.46 31.4 186. fv11 DES, so 34.55 t 185.9 352. t 45. MSIT 2.69 14.61 89. 164. rsat 33.1 0 124. 302.4 9 120. 16.6 61.6 169. 329. 3.8 ft.2 45.6 9 1800. 336. f 204. t que - 15.4 38.8 135. 306. rsA1 44. t 1800. 380. 9 160. Note: fras FSAR table 6.2.1-2 the absolute mariana values are Faaz a 48.1 ysis and Imax = 384.9 'T for all breaks taside containment. I
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- . .e Attachment 5, Appendix B.
Section II. ULNRC-1471 March 31,.1987
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t s CALLAWAY UPRATING SUBMITTAL. ATTACHMENT 5: APPENDICES APPENDIX B BOP - UPRATING LICENSING REPORT SECTION II - UNION ELECTRIC REVIEW f I y- si i i L _ _ _ . _ . _ . _ _ _ _ _ . _ _ _ _ _
e r I. INTRODUCTION The Callaway Plant is currently licensed to operate at a core power level of 3411 MWt which corresponds to a NSSS power level of 3425 MWt. Union Electric has proposed uprating the plant to a core power level of 3565 MWt which corresponds to a NSSS power 1.evel of 3579 MNt. All plant systems were screened ' for possible impact due to an uprating. Those systems that are-i clearly not dependent upon thermal power level were excluded from further engineering and licensing review. The results from a review of balance of plant systems performed by Bechtel are presented in Section I of Appendix B. Section II documents the results of a review performed by Union Electric on those balance of plant systems not included in the Bechtel review. Systems reviewed in Section II are listed in Table 1. II.
SUMMARY
AND CONCLUSION
~
The BOP systems that were reviewed by Union Electric include: circulating water; cooling tower makeup and blowdown; intake and water treatment; EHV switchyard bus; EEV switchyard 125 VDC; outgoing EUV lines; and the BOP computer. The Union Electric review also included the fuel pool cooling and cleanup system and the fuel building HVAC system. In conclusion, Union Electric's review of these secondary f ' side systems verifies that they have the capability to function properly at the uprated power level of 3579 !!Wt NSSS thermal power. The fuel pool cooling and cleaning system is not impacted by plant operation at uprated conditions, nor is the ability of ) the fuel building HVAC system impacted by plant operation at the j i increased power level, ~ The balance of plant systems reviewed can function properly
'without modification to their existing design. j III. SYSTEM EVALUATIONS i
circulating water System and cooling Tower The circulating water system, along with the condenser and cooling tower have been optimally designed to operate at the turbine-generator valves wide open (VWO) rating of 1234 MWe, 3562 MWt. The system and components are also capable of operating up to the uprated condition of 3579 MWt. The circulating water system is designed to supply cooling water at a constant rate of 530,000 gpm to condense steam while maintaining the turbine exhaust pressure below 5.0 in EgA. The uprated power level results in a greater water temperature rise and turbine exhaust pressure than experienced undt ; the turbine y
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generator guarantee rating. 'This results in losses in efficiency and power, as with.the guarantee rating (3425 MWt). The temper &ture of the water supplied to the condenser from the cooling tower is highly dependent on the weather conditions. The cooling tower is designed to supply 95 F water with a range of 78.6 F under weather conditions of 79 F WBT and 95 F DBT. However, weather conditions that result in water temperatures greater than 95 F do occur. Above approximately 97 degrees F water temperature, the 5.0 in EqA turbine exhaust back. pressure-However, this limit is reached and turbine load must be reduced. occurrence is infrequent and of short duration. The higher return water temperature due to the increased rating acts to increase the heat transfer capacity of the cooling tower, offsetting the greater heat rejection required, with the net result being a negligible change in cooling tower performance
- at the' higher rating.
l In addition to removing heat from the condenser during full
- power operation, the system must also remove heat when steam is directly bypassed to the condenser. Again, the system is designed to allow 40% of VWO main steam flow to the condenser at ]
/
the uprated conditions. Increased duty on the Service Water System (EA) from the uprating as reflected in slightly warmer service water returned to the circulating water system will have a negligible effect on the circulating water sytem. In summary, the circulating water system, cooling tower and condenser were orginally designed for the 3562 MWt rating and are operating within design at the guarantee rating. They are
- i- . expected to operate within design at the uprated condition of 3579 MWt. The difference in the pressures, temperatures, and j
~
*: flows listed in the FSAR for the 3562 MWt rating and the values j at the 3579 MWt rating are insignificant. i
)
Cooling Tower Makeup and Blowdown, Intake and Water Treatment The subject systems are all interrelated to provide adequate supplies of makeup water to the cooling tower at design ambient The temperatures and at the uprated condition of 3579 MWt. nakeup and i j intake and water treatment plant (WTP) and associated blowdown piping, with the exception of the clarifiers, were i designed for two unit operation and can provide up to 30,000 gpm with one intake pump out of service. The Callaway uprating will , increase current cooling tower evaporation rates by approximately l 4 percent, from 14,015 gpm to 14,575 gpm, total drif t and evaporation. Due to blow 0own rates required for radwaste dilution, there will be no additional lossec incurred for blowdown associated with uhe increase in evaporation rates. ; t
\
1 l L_ _ . - - .
.- O The WTP clarifier capacity had already been increased due to plant concerns of a clarifier outage during maximum evaporation and blowdown periods. The capacity of'each clarifier was modified and tested to provide water at the WTP at the design of Assu 15 ppm total suspended solids in the makeup water. increase of makming aeup water failure of one clarifier and a 4 percent associated with the uprating, the WTP has adequate capacity to i
provide for cooling tower makeup. Based on the above, there will be no adverse impact on plant operation of the subjecc systems due to the uprating. EHV Switchyard BUS, ERV Switchyard 125 VDC, Outgoing EEV Lines The review of these systems documented conformance to the safety design bases in the FSAR and conformance to Union Electric Company standards for EHV switchyards and EHV transmission lines.
*- The Callaway uprating will have no impact upon the ERV Switchyard 125 VDC system. The Union Electric transmission system is more
_- than adequate to carry the Callaway gross generation output of
- 1246 MWe (based on 3579 MWt rating) during both steady-state and transient conditions. A modest limitation on the Callaway Mvar
;~ output is required to avoid overloading the Callaway 1245 MVA Generator-Step-Up transformer. At full generation output of 1246 MWe with an assumed auxiliary transformer load of 53 MWe and 23 Mvar, the Callaway Mvar must be limited to 485 Mvar. This limitation, however, is not a problem from a system standpoint.
l The transmission system and the Callaway GSU transformer with the i Mvar limitations are more than adequate to accomodate operation of the Callaway Plant at uprated conditions. NSSS/ BOP Computers t~
- No hardware changes (such as wiring of additional inputs)
' are required for the NSSS and BOP computers. Software changes ,
will consist of the following: revised alarm limits for steam
- and feedwater flows and feedwater temperature revised constant I~ representing 100% power (used in calorimetric); and revised I constants associated with steam, feedwater, and RCS flow algorithms. These changes are consistent with changes made to control system setpoints.
Spent Fuel Pool Cooling And Cleanup System A reanalysis of the spent fuel pool (SFP) cooling system was performed to assure that discharged spent fuel assemblies would be adequately cooled and free from boiling for both normal and off-normal conditions. This reanalysis was performed to assess the impact of the SFP cooling system from the adoption of the VANTAGE 5 fuel design and plant operation at uprated conditions, including higher burnup levels.
1 s- ! The results of the reanalysis show that'the calculated peak j bulk pool temperatures for normal and off-normal condtions are 1 still within the existing TSAR limits. The calculated peak clad I temperatures are below the saturation temperature at the peak clad temperature. location. 1 l ' The : SFP cooling system will provide adequate cooling ' of , discharged spent fuel assemblies' to limit peak bulk pool j 3 I temperatures to below existing FSAR limits and to assure that the spent' fuel assemblies are free from boiling. The performance of the SFP cleanup system is not impacted by j l the adoption of the VANTAGE 5 fuel design or plant operation at l l uprated conditions. l \ 1 l Fuel Building HVAC A reanalysis was performed to assure that the plant operation at uprated conditions would not adversely impact the proper operation of the fuel building HVAC. i This analysis demonstrated that the fuel building EVAC is adequate to maintain an environment consistent with personnel comfort and safety. i The ability of the fuel building HVAC to limit the accidental release of radioisotopes to below hpplicable limits is f not impacted by. plant operation at uprated conditions.. - I i 4- -_.__--_m_ _ ._- _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ . . _
n- , .v.
-Table 1
.,-l _
Systems Reviewed for' Implementation Circulating." Water ' (DA):
. Cooling Tower' Makeup & Blowdown - (DB)
Intake' & Water- Treatment - (DE) . , Fuel. Pool Cooling'& Cleanup -(EC) s Fuel Building'HVAC'- (GG) EHV Switchyard Bus.- (MD)'
'"' EHV. Switchyard 125 VDC - (ME)
- ' Outgoing EMV Lines -- (MH)
BOP Computer'- (RJ) a 1 4 l-. (- 4 l _3_ f I' . - _ _ - - _ - - - _ _ - - _ _ - _ -
1- . 7 ,. ULNRC- 14 71' Attachment 6 Application Fee
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UNION ELECDUC COMPANY sr.um.nssoum - 219753 To cenTenac ersx or xtsserr.uissouai 4538 y CATE AMOUNT U 5 NUCLEAR REGULATORY 03 24 87 $ + * * * * *15 0 CD TOTHE ORDER Or COMMISSION > =toveat cou= Tensio =Arung i cestens ovra rs>. 1717 H ST NW THIS CHECK 5 II 5'd\ 8 d b / [' hdti;T, had
^; VOC AFTER oobus . .
[ l c==~"**' 4 e 2 &975 3n e a:OB & 50 2 4 59i: 84 035 & n' b I 1 L l H- ! u : 1 i I l l t._. l
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Env Etscruic V a 1901 Gratiot Street, St Lours Donald F. Schnell Vice President April 21, 1987
~
U.S. Nuclear. Regulatory Commission ATTN: Document Contr'ol Desk ,
/
Washington, D.C. 20555 Gentlemen: ULNRC-1494 CALLAWAY PLANT DOCKET NO. 50-483 CALLAWAY PLANT UPRATING SUBMITTAL
Reference:
ULNRC-1471, dated March 31,.1987 The referenced-letter transmitted an application for-
' amendment to the~Callaway. Plant Facility Operating License. The amendment request revises Technical Specifications to support.a-Callaway Plant uprating to the 3565 MWt' core power' level.
Included as an Attachment.to this letter.is Figure 9, titled, " Full DES at 102% Power, Failed MSIV Containment Temperature". This Figure was inadvertently omitted from Attachment 5, Appendix B,'Section.I of the referenced letter. It is hoped that the omission has not hampered any reviews. If there.are additional questions, please contact me. Very truly yours, , I
<7 j Donald F. Schnell 1 DJW/ dis Attachment j i
I
' Mamq Accress P O Box 149 S: Lou:s. MO 63166 = _ _ _ = _ _ _ . _ _ _ _ _ i
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C ITY ' OF . ST. . LOU I S ) Donald F. Schnell, of lawful' age, being first duly. sworn upon' oath says-that he .is Vice President-Nuclear and an officer of 4 : Union Electric Csmpany; that he has read the foregoing docum nt and knows the concont thereof; that he has executed the same for and on behalf of said company.with full power and authority;to do so; and ., that the facts therein stated are true'and correct t'o the best of his ' knowledge, .information and belief. - l By . D'ohald F. Sch'nell~
~
Vice President Nuclear day of ,:198 7 SUBSCRIBED and sworn'to before me this d/ t MS l BARBARA EPDF['a NOTARY PUBUC, $ TATE Of Mtss00Rt MY CONutS$10N EXPIRES APRIL 22, 1989 ST. LOUIS COUNTY
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a i l I 1 l.m 1 1 I E_ -' - - __. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ ______ g
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cc: Gerald Charnoff, Esq. Ghaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W. Washington, D.C. 20037 J. O. Cermak CFA Inc. 4 Professional Drive (Suite 110) Gaithersburg, MD 20879 W. L. Forney Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S. Nuclear _ Regulatory Commission RR#1 Steedman, Missouri 65077 Tom Alexion (2) Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 316 7970 Norfolk Avenue Bethesda, MD 20014 Ron Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 i l i s~_____--_ _ _ '}}