ML20052E799
| ML20052E799 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/22/1979 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20052E790 | List: |
| References | |
| 18-1005812, 18-1005812-00, NUDOCS 8205110431 | |
| Download: ML20052E799 (34) | |
Text
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BWP-20004 (6-76)
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, _ BABCOCK & WILCOX NUCLEAR POwit GENfRATION Divl510N TECHNICAL DOCUMENT FllNCTIONAL CONTRACT SPECIFICATION (r -
18 - 2005812 ._00 Doc. ID - Serial No., Revision No.
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for REACTOR C00LAhT SYSTEM e
e PAGE 1 8205110431 820504 PDR ADOCK 05000302 P PDR
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ACCEPTABILITY OF FUNCTIONAL SPECIFICATION 18-1005812-00 CERTIFICATION DOCUMENT
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Babcock & Wilcox Co. Contract No. 620'-0007-50 h.
j, User: Florida Power Corporation ,
User Contract No. PR3-1000 I certify that Functional Contract Specification 18-1005812-00 has been reviewed to show that the primary piping, as shown by the Babcock
& Wilcox Co. drawings identified on pages iv and v of the Stress Report Certification Document dated January 27, 1976, would meet the design requirements of Tentative American Standard Code for Pressure Piping, Nuclear Power Piping U.S. A.S. B31.7, February 1968. The basis of this 4E - review and acceptability of the Functional Specification for a licensing rs. Power upgrade of this contract consisted of both a comparison of the 44;_ specification with the corresponding Functional' Specification of a i.* physically duplicate component, B&W Contract 620-0012-50, and an assump-tion that the stress evaluation of B&W Contract 620-0012-50 provides accurate assurance of acceptable code stress values. With the above conditions, the component would be shown to meet applicable code stress limits if reanalyzed.
Exceptions:
(1) The stress limit condition of Table 3, Case iII, of the m Functional Contract Specification 18-1005812-00 is M excluded from this certification.
g This certification consists of one (l) sheet. Attested to this date 2-2-79.
By: 6N N WJ License No.
15108
, Coburn 0. Jacob, Sfnior Engineer Indiana State Board Component Engineefing of Professional Mt. Vernon Engineering Dept. Engineers
&p- The Babcock & Wilcox Company
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CERTIFICATI0fi DOCU?'EllT FOR REACTOR VESSEL ACCEPTABILITY OF -
. FullCTI0?1AL SPECIFICATION 18-1005812 LOO -
SHEET 10F1 Babcock 5WilcoxContractflo. 620-0007-51/52 User: Florida Power Corp.
User Contract !!o. PR3-1000 The reactor vessel stress report last certified by N.G. Dadiras on 3-27-74 has been reviewed and is acceptable to functional specifica-tion 18-1005812-00 with the following notation:
Loading condition III in table 3 of specificatien 18-1005812-00, which was included in the original functional specification has not been analysed nor was it required by the specifications or codes.
Attested to this date 2-2-79.
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By: )<W/44(/. Ne'~1 License flo. 16679 Uouglas A. Huston Indiana State Board Mt. Vernon Component of Professional Engineers Engineering ,
The Babcock & Wilcox Co.
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( * *: No.16679 i
( . STATc OF
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ACCEPTABILITY E FUNCTIONAL SPECIFICATION 18-1005812-00 CERTIFICATION DOCUMENT d'
Babcock & Wilcox Co. Contract No. 620-0007-59 3* '
p User: Florida Power Corporation - -
User Contract No. PR3-1000 I certify that Functional Contract Specification 18-1005812-00 has been reviewed to show that the Pressurizer, as shown by the Babcock &
Wilcox Co. drawings identified on page 3 of the Stress Report Certifica-tion Document dated February 18, 1974, would meet the design require-ments of the ASME Code,Section III,1965 Edition with Addenda through Sumer 1967.
pp .
g The basis of this review and acceptability of the Functional W Specification for a licensing power upgrade of this contract consisted of both a comparison of the specifiestion with the corresponding Functional Specification of a physically duplicate component, B&W Con-tract 620-0012-59, and an assumption that the stress evaluation of B&W Contract 620-0012-59 provides accurate assurance of acceptable code tr:
(
stress values. With the above conditions, the component would be shown to meet applicable code stress limits if reanalyzed.
Exceptions:
t.sss-d (1) The stress limit condition of Table 3, Case III, of the Func-tional Contract Specification 18-1005812-00 is excluded from this certification.
K lM k This certification consists of one (1) sheet. Attested to this date 2-2-79.
, anr l .Ei By: @. Ld. Sele /M License No. 17881
,k R.W. Schaf fst' din, Associate Engineer Indiana State Board IN"- Component Engineering of Professional 74 Mt. Vernon Engineering Dept. Engineers The Babcock & Wilcox Company -
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ACCEPTABILITY CF FUNCTIONAL SPECIFICATION 18-1005812-00 C CERTIFICATION DOCUMENT Babcock & Wilcox Co. Contract No. 620'-0007-55 User: Florida Power Corporation .
User Contract No. PR3-1000 I certify that Functional Contract Specification 18-1005812-00 has been reviewed to show that the Once Through Steam Generator, as shown by the Babcock & Wilcox Co. drawings identified on page 4 and 5 of the Stress
, Report Certification Document dated 3/27/74, would meet the design require-i ments of ASME Code,Section III,1965 Edition with Addenda through Summer 1967. The basis of this review and acceptability of the Functional Specification for a licensing power upgrade of this contract consisted of both a comparison of the specification with the corresponding Functional Specification of a physically duplicate component, B&W Contract
- 620-0012-55, and an assumption that the stress evaluation of B&W Contract
! 620-0012-55 provides accurate assurance of acceptable code stress values.
With the above conditions, the component would be shown to meet applicable code stress limits if reanalyzed.
( Exceptions:
(1) The stress limit condition of Table 3, Case III, of the Functional Contract Specification 18-1005812-00 is excluded from this certi-fication.
! This certification consists of one (1) sheet. Attested to this date 2-2-79.
t By: ko_l d M w nin License No. 14730 Ralph' Skinner, Senior Engineer Indiana State Board Component Engineering . of Professional Mt. Vernon Engineering Dept.
The Babcock & Wilcox Company
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4 s s 'THE~ BABCOCK & WILCOX COMPANY
' POWER GENERATION GROUP Ta l w..u:,g g M (5 - PROJECT MANAGER From J. M. BURNETT - RCS COMPONENTS m 66u Cust. File No A20-0007-T3.55/
FLORIDA POWER CORPORATION or Ref. T3.59/T3.51/T3.5 Subj. Date POWER UPGRADE - STRESS REPORT REVIEW FEBRUARY 20, 1979 Attachments: (1) Stress Report Certification Document for Steam Generator, certified by Ralph Skinner on 2/2/79.
(2) Stress Report Certification Document for RC Piping, certified by Coburn O. Jacob on 2/2/79.
(3) Stress Report Certification Document for Reactor Vessel, certified by Douglas O. Houston on 2/2/79.
(4) Stress Report Certification Document for Pres-surizer, certified by R. W. Schaffstein on 2/2/79.
(5) Functional Contract Specification for Reactor Coolant System, Document # 18-1005812-00.
The original stress analyses for the Steam Generator, RC Piping, Pressurizer, and Reactor Vessel have been reviewed to assess their applicability in light of revised functional requirements contained in Attachment (5). This review was done by means of comparison of the revised FPCo requirements w.th the corresponding requirements and stress analyses of a physically duplicate set of components on another B&W contract. No unique reanalysis was done for FPCo.
The review performed indicates that the FPCo components would meet the requirements of the original contract Codes if reanalyzed. A copy of each component recertification (Ittachment ' 4 is attached.
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j J M. Burnett JMB:dla cc: R.M. Douglass T.C. Helms D.C. Leinhart (w/att. 1-4 only) g yd
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R.B. Park eM-R.N. Tornow C. E. Enksdale l
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DOCUMENT gSUBut '.
LFORM
( Babcock & Wilcox # @%b Po , c.n.,.uon c, u,
- g. , P.O. Box 1260. Lynenburg, Va. 2.:505 g ', :, % {,<
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.5 Telephone: (804) 384 5111 A '
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TO Florida Pouer Corooration 2/22/79 Crystal River, Florida N'M B&W CONTR ACT NO.
CUST Florida' Power Corooration CUST. ORDER NO SHEET 1 OF 5 TYPE DOC c fTN Mr. G. P. Beattv 1 Copy ea. O uNDER SEP COVER O ENCLOSED D FOR COMMENTS & APPROV AL SY O UNDER SEP COVER O ENCLOSED D FOR INFORM ATI ON ONLY REI ERENCES O HAS BEEN REV1 SED AS PER YOUR PREVIOUS COMMEN TS OF O rOR rlN AL DISTRIBUTION O FURTHER EXPL AN ATION REQUI RED SEE BELOW OR ATTACHMENTS X Quality Document Submittal DOCUMENT DESCRIPTION l
i VENDOR B& W COMP GRP DOC.
ORIGIN DOC.NO. DO C . NO . NO. NO. TITLE B&W 18-1005812-00 - -
Functional Contract Specification
! As a result of the CR-3 power upgrade to 2544 Wt a new specification was prepared to esta-blish the General Functional Requirements for reactor coolant system components. Enclosed, Plance find one copy of the functional specification and certifications of compliance for the reactor vessel, piping, stea= generators, and pressurizer which concludes that the CR-3 c:mponents will meet the design requirements of the functional specification.
C2rtificates of compliance should be filed in your Quality Assurance folders for the respective l crmponents and the functional specification should be maintained in your Quality Assurance foldsrs which contain all NSS and Auxiliary System Component Specifications.
Ctpics of this transmittal letter have been forwarded to personnel within the FPC Licensing S:ction in the event reference need to be made to the NRC during the licensing process.
If ycu have any questions or comments, please advise.
l
- 'd suA M "C .-E. Barksdale c Q. B. DuBors w/o att. J. S. Laing w/o att.
E. C. Simpson w/o att. J. R. Shetler w/o att. bec: R. J. Finnin w/o at P. Y. Baynard w/o att. W. P. Ellsberry w/o att. J. M. Burnett w/o at NSC-12D-12B w/o at The Babcock & Wilcox Company / Established 1867
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BWNP-20007 (6-76)
BABCOCK & WILCOX m ,,
NUCtfAt Powlt otNftAfloN DIVI $loN C TECHNICAL DOCUMENT 28-2005812-00 6.17.3 Transient Data 6.17.3.1 Reactor Coolant System The reactor coolant pressure, . temperature and surge line flow rate are shown in Figure 16-1. The make-up and high pressure injection flow rate and tempera-ture are shown in Figure 16-2. In this transient, the makeup alone is not able to supply enough water to makeup for the contraction losses of the reactor coolant system and the low pressure causes the high pres-sure injection system to be turned on. De flow shown is split between the 4 HP injection no::les, one of which has been kept cool by the makeup flow.
De other three are at reactor coolant temperature when the cold water from the borated water storage tank begins being injected into the reactor coolant system.
6.17.3.2 Steam Generators The steam generator conditions for the steam line break are presented in Figure 16-3. The blowdown thrust
_ forces are defined in Reference 3 of Appendix II.
6.18 Transient 17 ASB (Upset 6 Emergency Conditions) 6.18.1 General Description Transients 17A and 17B are to be used for analysis of uausual operations of the steam generators. Transient 17A is a tran-sient in which feedwater flow is lost to a steam generator, which causes a reactor trip, and the steam generator is evap-orated to a dry, pressurized condition. Following the reactor trip, auxiliary feedwater is slowly introduced to the dry steam generator to obtain minimum level. Transient 17B is an emergency transient in which a turbine bypass valve is assumed to stick cpen.
The affected steam generator blows down to a dry, depressurized I condition, and a reactor trip occurs. The faulted bypass valve is isolated and feedwater is slowly introduced through the auxiliary feed no::les on the dry steam generator until l minimum water level and pressure are restored.
6.18.2 Cycles The number of loss of feedwater to one steam generator events for design purposes shall be 20. This event is included in the reactor trip cycle category and a complete cycle will consist l
of: a) Loss of feedwater to one steam generator with resultant reactor trip and recovery of level in the steam generator, and b) Return to full power as described in Transient 3 (Power Loading) .
DATE: PA E 29 10-6-78
BWNP-20007 (6-76)
BABCOCK & WILCOX wumeen NUCLIAt Powie GENenAfloN Divi 5 ton C TECHNICAL DOCUMENT I 28-2 "5822- "
N The number of stuck-open bypass valve events for design purposes shall be 10. This event is included in the reactor trip cycle category and a complete cycle will consist of: a) Stuck-open bypass valve with reactor trip and resultant depressuri-
- ation of the steam generator and recovery of steam generator level, and b) Return to full power as described in Transient 3 (Power Loading).
6.18.3 Transient Data 6.18.3,1 Reactor Coolant System The reactor coolant system conditions for Transient 17A are presented in Figures 17A-1, 17A-2, and 17A-3.
Figures 17A-1,17A-2, and 17A-3 show the conditions during loss of feedvater flow and Figures 17A-4,17A-5 and 17A-6 show conditions during restart of the steam gener-ator. Restart of the steam generator is accomplished with: the auxiliary feedwater no::le.
The reactor coolant system conditions for Transient 17B are presented in Figures 17B-1,17B-2, and 17B-3.
These figures show the conditions during depressuri-
- ation and Figures 17B-4 and 17B-5 show the conditions during restart of the steam generator.
6.18.3.2 Steam Generator l The steam generator conditions for Transient 17A are l
presented in Figures 17A-4,17A-5 and 17A-6.
The steam generator conditions for Transient 17B are presented in Figures 17B-4 and 17B-5.
6.19 Transient 18 - Loss of Feedwater Heater (Doset Condition) 6.19.1 General Description The feedwater heaters are arranged so that one or more heaters can be isolated if necessary. When a feedwater heater is iso-lated, the feedwater temperature entering the steam generators will decrease. The control system will cause feedwater flow changes necessary to hold reactor power constant and no pertur-bations are expected in the reactor coolant system.
DATE:
10-6-78 30 1
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c 5000 0 1C 2C 3C 4C SC 60 7C BC 9C 10C Time, Seconcs FIGURE 16-1
' TRAN3IENT Nt 16 (STEAN LINE FAILURE)
( REATOF CJLANT 513fEW PARANETER$
DATE: 10-6-78 SERIAL: 18-1005812-00 I
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O 'l 2 3 4 5 6 7 8 9 10 11 12 13 Time, Minutes FIGURE 16 2 TRANSIENT NO. 16 (STEAN LINE FAILURE) l NAKEUP AND HP INJ ECTION FL0n R ATE AND TEWPERATURE DATE: 10-6-78 SERIAL: 18-1005812-00 i
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FEEDIATER-STEAR SYSTEF PARAMETERS i
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DATE: 10-6-78 SERIAL: 18-1005812-00
Reactor 602.8
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Inlet l l Temperature,*F l l l 1 2700 2450 - l Reactor 2200 Pressure, psia 1950 1700 l l l 0 1 2 3 4 5 6 Time, Winutes l
FIGURE 17A-1 TRANSIENT NO. 17A (LOSS OF FEE 0 TATER TO ONE STE AW GENER ATOR) RE ACTOR COOL ANT TEMPER ATURE AND PRESSURE
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DATE: 10-6-78 SERIAL 16-1005S12-00
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l DATE: 10-6-78 SERIAg: 18 1005S12-00 i
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TRANSIENT NO 17A (LCSS OF FEED ATER FLD TC ONE STIAN GENERATOR)sAKEUP ANC $ PRAY TEWPERA- .
TURE AND FLot RATE DATE': 10-6-78 SERIAL: 18-1005812-00 1
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Temperature, F 200 17A E (ALSO FR TEMP" 100 f AUI. & MAIN O N0Z2LES OF ORT SG -
! UNTIL RESTART. ORT SG Ft N0ZZLES BC *PPRDACH SHELL -
Degrees of 60 ' TEMP.) .
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TRANSIENT NO. 17A (LOSS OF FEE 0 TATER TO ONE STEAF GENERATOR) STEAN FEEDRATER SYSTEW PARAMETERS DATE: 10-6-78 SERIAL: 18-1005812-00
570
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O t 2 3 4 5 Time Winutes after Restart l FIGURE 17A 5 TRAN5lENT NO.17A (LDSS OF FEE 0eATES 70 Oht STEAP GEhERATOR
. ST ARTUP OF OR Y" PRES 3URIZED STE ar CINER ATOR l
DATE: 10-6-78 SERIAL: 18-1005812-00
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1450-4 6 6 10 12 0 2 Ties, Ninutes FIGURE 17E-1 l
, TRANSIENT 17E (STUCK DPEN TUREINE EYPASS VALVE) l REACTOR COOLANT TEMPERATURE AND PRESSURE DATE: 10-6-78 SERIAL: 18-1005812-00 y -- - . ,
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4 000 0 2 e s i Time, sinutes FIGUR[ 171 2 IRAN!![NT NC 175 (STUCK DP[>
- TURtlN[ IfPA!! VALY!) SURG [ LIN[
WAT[a T[WP[RATUR! AND FLDS l DATE: 10-6-78 SERIAL: 18-2005812-00
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(
FIEURE 17B 3
! TRAN5tlNT NO. 170 (STU:K OPEh TURBih! BTP ASS VALVE )
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r DATE: 10-6-78 SERIAL: 18-1005812-00 l
100 j Steam BC ^O I I I I i l i F len. S SC 1i I I I I I gg
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,C i I \ M f 100 C
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0 1 2 3 4 6 6 7 e Time. Minutes FIGURE 175-4 TRANStENT NC. 178. (STUCF, OPEh TURBINE BTPASS VALVE )
STEAN FEEDIATER SYSTEN PARAMETERS DATE: 10-6-7S SERIAL: 18-1005812-00 l
-e 600 '
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Time, Minutes af ter Restart FIGURE 178 5 TRANSIENT NO. 178 (STUCK OPEN TURBINE BYPASS V ALVE ) STARTUP OF " DRY" DEPRESSURIZIO STEAN GENERATOR
~
l DATE: 10-6-78 SERIAL:
~
l 18-1005812-00 l
t
- - - - - - ~ . ~ . . .
BWNP-20007 (6-76)
BABCOCK & NVILCOX NUA4BER NUCLEAR PoWit ofNERAfloN DIVI $loN TECHNICAL DOCUMENT 18-2005s12-00
- 5) Load reduction to hot standby following load rejection.
The maximu= loading or unloading rates are : 3/41./ min.
6.3.2 Number of Cycles The total number of cycles for design purposes shall be 1440.
6.3.3 Transient Description - Ot to 15% Power 6.3.3.1 Reactor toolant System The reactor coolant system conditions are shown in Figures 2A-1, 2A-4, and 2A-5.
The pressuri:er water inventory is shown on Figure 2A-1 for letdown rates of 70 and 140 gpm.
6.3.3.2 Steam Generators The steam generator conditions are shown in Figures 2A-2 and 2A-3.
The delay in feedwater temperature shown on Figure C 2A-3 results from the flow time between the first point feedwater heater and the steam generator. The steam and feedwater flows required at 150 power increase as the feedwater temperature increases.
6.3.4 Transient Description - 15% to 0% Power 6.3.4.1 Reactor Coolant System The reactor coolant conditions are shown in Figures 2B-1 and 2B-4 The pressurizer inventory shown for this transient is based on starting from normal water level, and 108 gpm total makeup flow is required to maintain the indicated inventory.
I i 6.3.4.2 Steam Generators The steam generator conditions are shown in i Figures 2B-2 and 2B-3.
1 The feedwater temperature delay results from the flow delay between the first point feedwater heater and the steam generator. The steam and feedwater flow require-ments at a given power level decrease as feedwater
(
temperature decreases.
l DATE: PAGE 10-6-78 16
BWNP-20007 (6-76)
BABCOCK & WILCOX . . , ,
NuCLEAA Powet otNteAfsoN OfVl$ son C. TECHNICAL DOCUMENT 18-2005812-00 Hot standby is defined as operation at power levels equivalent to decay heat level (0 to 3% thermal power) plus heat input from up to four reactor coolant pumps.
During this operation the feedwater flow is routed through the main feedwater no::les. If there are perturbations in feedwater flow a themal transient could occur in the main feedwater no::les although a feedwater temperature is constant. A typical cyclic feed-water flow t?s '- is shown in Figure 2B-5.
The number of potential cycles for the main feedwater no::les in hot standby is 120,000, 6.4 Transient 3 - Power Loading 8% to 100's Power (Normal Condition) 6.4.1 General Description .
This transient is the design power loading cycle. Starting at 8% power the reactor pouer is manually increased to 15% power at rates up to 3/4'4/ min, placed in automatic control and power is
- then increased at rates up to 5%/ min between 15% and 20% power,
( 10%/ min between 20% and 90'6 power and 5%/ min between 90% and 100%
( powe'. r The transient curves are based on loading at 3/4%/ min up to 15% power, and 10%/ min from 15% to 100's power.
6.4.2 Loading Cycles I The total number of loading events for design purposes shall be 48,000. This corresponds to approximately 3 full loadings per day for 40 years plant life.
{
l 6.4.3 Transient Data 6.4.3.1 Reactor Coolant System l The reactor c!oolant system conditions for power loading are shown on Figures 3-1, 3-2, and 3-3.
The initial water inventory at the start of the transient is assumed to be at minimum water level, 350 ft3 The letdown rate is assumed to be constant and the pressurizer
, inventory rises with temperature during the first part of
( the transient.
6.4.3.2 Steam Generators The steam generator conditions for power loading are shown in Figure 3-4 The maximum and nomally expected steam temperatures are shown.
DATE: PAGE 10-6-78 17
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DATE: 10-6-78 SERIAL: 18-1005812-00 n --
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