ML20083J697: Difference between revisions
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| number = ML20083J697 | | number = ML20083J697 | ||
| issue date = 12/30/1983 | | issue date = 12/30/1983 | ||
| title = Forwards Applicant | | title = Forwards Applicant Re Changes to Byron Steam Generators to Minimize flow-induced Vibration.Ltr Addresses Issues Litigated Under Intervenor Contentions Challenging Integrity of Installed Steam Generator Tubes | ||
| author name = Gallo J | | author name = Gallo J | ||
| author affiliation = CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.),, ISHAM, LINCOLN & BEALE | | author affiliation = CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.),, ISHAM, LINCOLN & BEALE | ||
| Line 11: | Line 11: | ||
| contact person = | | contact person = | ||
| document report number = NUDOCS 8401050366 | | document report number = NUDOCS 8401050366 | ||
| title reference date = 12-16-1983 | |||
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, LEGAL/LAW FIRM TO NRC | | document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, LEGAL/LAW FIRM TO NRC | ||
| page count = 28 | | page count = 28 | ||
| Line 21: | Line 22: | ||
* '' ~ | * '' ~ | ||
, TELEPHONE 312 558 7500 TELD 2-5288 December 30, 1983 Ivan W. Smith, Esquire Dr. Richard F. Cole Administrative Judge and Administrative Judge Chairman Atomic Safety and Licensing Atomic Safety and Licensing Board Board U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. A. Dixon Callihan Administrative Judge Atomic-Safety and Licensing Board c/o Union Carbide Corporation P. O. Box Y Oak Ridge, Tennessee 37830 Gentlemen: | , TELEPHONE 312 558 7500 TELD 2-5288 December 30, 1983 Ivan W. Smith, Esquire Dr. Richard F. Cole Administrative Judge and Administrative Judge Chairman Atomic Safety and Licensing Atomic Safety and Licensing Board Board U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. A. Dixon Callihan Administrative Judge Atomic-Safety and Licensing Board c/o Union Carbide Corporation P. O. Box Y Oak Ridge, Tennessee 37830 Gentlemen: | ||
I am enclosing a letter dated December 16, 1983 from Mr. Tramm of Commonwealth Edison Company to Mr. Harold R. Denton. | I am enclosing a {{letter dated|date=December 16, 1983|text=letter dated December 16, 1983}} from Mr. Tramm of Commonwealth Edison Company to Mr. Harold R. Denton. | ||
This letter encloses information concerning the changes being made to the Byron steam generators to mininize flow-induced vibration. This matter was one of a number of issues litigated under the Intervenors' contentions challenging the integrity of the steam generator tubes installed at the Byron Station. | This letter encloses information concerning the changes being made to the Byron steam generators to mininize flow-induced vibration. This matter was one of a number of issues litigated under the Intervenors' contentions challenging the integrity of the steam generator tubes installed at the Byron Station. | ||
The information covers certain plant-specific data requested by the NRC Staff. It does not change in any way the testimony filed by Commonwealth Edison Company on the steam generator tube integrity contentions. | The information covers certain plant-specific data requested by the NRC Staff. It does not change in any way the testimony filed by Commonwealth Edison Company on the steam generator tube integrity contentions. | ||
| Line 37: | Line 38: | ||
==Subject:== | ==Subject:== | ||
Byron Generating Station Units 1 and 2 Braidwood Generating Station Units 1 and 2 Steam Generator Tuoe Vibration NRC Docket Nos. 50-454, 50-455, 50-456, and 50-457 References (a): February 9, 1983 letter from T. P. Tramm to H. R. Denton. | Byron Generating Station Units 1 and 2 Braidwood Generating Station Units 1 and 2 Steam Generator Tuoe Vibration NRC Docket Nos. 50-454, 50-455, 50-456, and 50-457 References (a): {{letter dated|date=February 9, 1983|text=February 9, 1983 letter}} from T. P. Tramm to H. R. Denton. | ||
(b): April 12, 1983 letter from T. R. Tramm to H. R. Denton. | (b): {{letter dated|date=April 12, 1983|text=April 12, 1983 letter}} from T. R. Tramm to H. R. Denton. | ||
(c): July 18, 1983 NRC Summary o f Meeting on July 7, 1983 with the Technical Review Committee. | (c): July 18, 1983 NRC Summary o f Meeting on July 7, 1983 with the Technical Review Committee. | ||
(d): July 18, 1983 letter from L. D. | (d): {{letter dated|date=July 18, 1983|text=July 18, 1983 letter}} from L. D. | ||
Butterfield, Jr. to D. G. Eisenhut. | Butterfield, Jr. to D. G. Eisenhut. | ||
(e): August 1, 1983, letter from L. D. | (e): {{letter dated|date=August 1, 1983|text=August 1, 1983, letter}} from L. D. | ||
Butterfield, Jr. to D. G. Eisenhut. | Butterfield, Jr. to D. G. Eisenhut. | ||
Latest revision as of 04:27, 26 September 2022
| ML20083J697 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 12/30/1983 |
| From: | Gallo J CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), ISHAM, LINCOLN & BEALE |
| To: | Callihan A, Cole R, Smith I Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8401050366 | |
| Download: ML20083J697 (28) | |
Text
{{#Wiki_filter:.__ _ __ ____ S' f j ISHAM, LINCOLN & BEALE DOCKET,ED COUNSELORS AT LAW Ub b 1120 CONNECTICUT AVENUE. N W. = SU!TE 840 - EDWARD S.1$ HAM, 1872-1902 WASHINGTON. D C. 20036 202 833 9730 g [; ==) h* CHICAGO OFFICE RO8ERT T. UNCOLN. 1872-1889 THRE E FIRST NATIONAL PLAZA WILLIAU G. BEALE. 1885-1923 CHICAGO. ILDNOIS 60602
* ~
, TELEPHONE 312 558 7500 TELD 2-5288 December 30, 1983 Ivan W. Smith, Esquire Dr. Richard F. Cole Administrative Judge and Administrative Judge Chairman Atomic Safety and Licensing Atomic Safety and Licensing Board Board U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. A. Dixon Callihan Administrative Judge Atomic-Safety and Licensing Board c/o Union Carbide Corporation P. O. Box Y Oak Ridge, Tennessee 37830 Gentlemen:
I am enclosing a letter dated December 16, 1983 from Mr. Tramm of Commonwealth Edison Company to Mr. Harold R. Denton. This letter encloses information concerning the changes being made to the Byron steam generators to mininize flow-induced vibration. This matter was one of a number of issues litigated under the Intervenors' contentions challenging the integrity of the steam generator tubes installed at the Byron Station. The information covers certain plant-specific data requested by the NRC Staff. It does not change in any way the testimony filed by Commonwealth Edison Company on the steam generator tube integrity contentions. Sincerely, i
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6 Jose h Gallo cc: Service List , 8401050366 831230 PDR ADOCK 05000454 G PDR
k / N C mmonwr_Ith Edison
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. ( O 7 Address Reply to Post Office Box 767 (j Chrcago litenois 60690 00CMETED USNRC December 16, 1983 ..
'84 J3 -3 A10 :19 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Generating Station Units 1 and 2 Braidwood Generating Station Units 1 and 2 Steam Generator Tuoe Vibration NRC Docket Nos. 50-454, 50-455, 50-456, and 50-457 References (a): February 9, 1983 letter from T. P. Tramm to H. R. Denton. (b): April 12, 1983 letter from T. R. Tramm to H. R. Denton. (c): July 18, 1983 NRC Summary o f Meeting on July 7, 1983 with the Technical Review Committee. (d): July 18, 1983 letter from L. D. Butterfield, Jr. to D. G. Eisenhut. (e): August 1, 1983, letter from L. D. Butterfield, Jr. to D. G. Eisenhut.
Dear Mr. Denton:
This letter provides additional information regarding the changes being made to the Byron and Braidwood steam generators to minimize tube vibration. NRC review of this information should enable closure of Outstanding Item 10 of the Byron SER. As described in previous generic correspondence meetings and hearings, selected steam generator tubes are being expanded at two tube support plates and 10% of the main feedwater flow is being diverted to the aur.111ary feedwater nozzle. Extensive reviews and analyses of these modificetions have been performed by Westinghouse and by the Counterflow Steam Generator Owners Review Group. These efforts have verified that the modifications will be ef fective in reducing tube vibration to accept-able levels and that the modifications will introduce no unacceptable safety consequences during normal or transient operating conditions.
H. R. Denton December 16, 1983 This letter provides the plant-specific information, details of this modification and the results of plant-specific safety analyses on the split-feed arrangement. Tables 2.5-1 and 2.5-2 of references (d) and (e) list the FSAR sections and design transients which have been reviewed. Special attention has been given to the prevention of waterhammer and to the impact of these changes upon plant operating procedures. We have concluded that the Byron and Braidwood plants can be operated safely with the modified steam generators and feedwater piping. Attachment A to this letter contains revised FSAR pages. These contain the necessary revisions to the facility description and accident analyses. These pages will be incorporated into the FSAA at the earliest opportunity. As indicated in references (d) and (e), inservice inspection plans and tube plugging criteria are also being addressed on a plant-specfic basis. For Byron and Braidwood, the Technical Specifications already proposed adequately cover these issues. The extensive inservice inspection program already agreed upon provides for the early detection of unanticipated problems with steam generator tubing. The 40% plugging criteria appears adequate to prevent mid-cycle fallure of any tubes which are found to experience minor degradation. To provide additional assur-ance of the adequacy of the tube modifications, vibration measurements are to be made on selected tubes during operation of one of the first domestic units with expanded tubes. If these measurements indicate the need for additional inservice inspection, ISI changes can be easily incorporated.into individual plant Technical Specifications such as Byron's. Please address further questions regarding this matter to this One signed original and fifteen copies of this letter and the enclosures are provided for NRC review. Very truly yours, [ b, 'ett . T. R. Tramm Nuclear Licensing Administrator 1m Enclosure 7379N
ATTACHMENT A-FSAR Revisions to Implement 90/10 Feedwater Flow Split
- 1. Table 3.6-12 Corrections
- 2. Table 3.9-16 Revisions
- 3. Table 4.1-1 Revisions and Corre:tions
- 4. Table 4.4-1 Revisions and Correct ions
- 5. Figure 4.4-9 Revisions
- 6. Page 5.1-4 Revisions
- 7. Table 5.1-1 Revisions
- 8. Figure 5.1-2 Revisions
- 9. Section 5.4.2.5.3 Replacement Section 10.
Section 5.4.2.5.4 New Section
- 11. Table 6.2-58 Revisions
- 4
- 12. Section 10.4.7.3 Replacement Section
- 13. Section 15.0.3.2 Revisions
- 14. Figure 3.6-2 Revisions
- 15. Figure 10.4-1 Revisions
- 7379N
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TABLE 3.6-12 CALCULATED STRESSES FOR POSTULATED BREAK POINTS (For ASME Sec. III Class 2&3 and ANSI B31.1 Piping Systems) PIPING SYSTEM CALCULATED STRESS NORMAL & UPSET BREAK PLANT CONDITIONS LINE NUMBER (S) ID
- 0. 8 (1. 2Sb +SA)
(psi) (psi) FEEDWATER Loop 1 1FWO3DA-16" B20A 17660 32400. 1FWO3DA-16" B20B 16832 32400. ." 1FWO3 DA-16 " B40A 13586 32400 f 1FWO3DA-16" B65A 19150 R in 1Fw03DA-16" B65B 32400 y
- 20934 32400 N 1FWO3DA-16" B80 15907 32400 $
- n FEEDWATER Loop 2 1FWO3DB-16" BSA 10706. 32400 1FWO3DB-16" B30A 17847. 32400 1FWO3DB-16" B30B 17293. 32400 1FWO3DB-16" B55A 12908. 32400 1 FWO 3 DB-16 " B85A 16971. 32400 1FWO3DB-16" B85B 18565. 32400 1FWO3DB-16" B100 14974. 32400
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n/u-rBAR TAuLU 3.9-It, (Con t ' til ACTUATED SIZE DODY QUALITY ISI TAG NUMBER BY (in.) TYPE GROUP CATEGOltY l'a l D ICV 8378A,B -- 3 Check A C M-64-5 1CV9379A,B -- 3 Check A C M-64-5 1CV8381 -- 3 Check 's r M -(. 4 - 5 1CV8442 -- 2 Chuck Il C M-84-4 1CV8481A,B -- 4 Chuck 11 C M-64-1 1CV8546 -- 8 Check in C M- #.4 -4 1CV8804A Motor U Cato 16 It M-h4-4 1FC009 -- 4 l'luy 11 A M - t. 'l - 1 IFC010 -- 4 l'l uy si A M-6 l- 1 1PP010 Air 4 Gato n n M *2-t 1FP0ll - Air 4 Catu li n M-52-1 1rWOO9A-D llydr nul it? I f, da t.o n n M t t. t l 1rwolbA-Il -- .75 u l ot,o I. n M 11. I 1rWO35A-D Air J GaLu at it M - n. I 1FWO36A-D -- 3 Check to e M-16-1 Irwo19A-p Air t. um t is n n M i.. i i I P'WU 41 A-li Ali i ti t olio n n M i s. I IFWO75A-D -- 6 Check 15 C M-16-1 l i 01 1 . n a M ' 1 I AOM Ale
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T ABLE 4.1-1 s PEAC70P DESIGN CDMPARISON TABLE i 4 BYRON AND 3RAIDWOOD BTROE AMC BRAIDWOOD i UNITS I and 2 WCAP-958@ UNITS I and 2 (tant PARASITIC PUELL lOPTIw! ZED FCEL) . JEFr2ME CEgg THERMAI. AND HYDRA 0LIC DESIGN PmwETERS 3'II 381I Feactor Core Heat Output, (1004), MW t 3811
- 1. 11641.7 11641.7 Reactor Core Heat Output, 100 Sta/hr 11641.7
- 2. 97.4 97.4 Heat Generated in Fuel, t 97.4
- 3. 2280 22!3 '
Core Pressure, Nominal, psia II 225* I 4. II 222: 225C 225"
- 5. System Pressare, Minimus Steady State, psia
- 6. Minimur DNBR at Nominal Conditions 2.4* 2.40 2.05 Typical Flow Channel .12 2.!f 1.74 Tt.imble (Cold Wall! Flow Channel winimum DNBR f or Design Transients
,. 7.
>1.3* >l.49 >;.44-7 Typical Flow Channel -
>1.4' >;.4~
* >1.3 Thimble Flow Channel WI3-; 2) WI3 ;I2'
*R* (W- 3 w . t h mod i f i ed
- 8. DNB Correlation Spacer Fa: tor)
E9EbAST FLUK 0 '. 4 4 . - 142 l Total Trersal Flow Rate, 10 lb,'tr 138.E " 9. A
- 10. Effective riow Rate fo! feat Tra-ster, ,);,. g
, ;, t .
106 3 p g ,h, 54.; 54..
- 11. E'fective Flow Area for Seat Transfer, ft 51.1 If.! '!.-
- 12. Aver age Velocity Along F;el Rods, f t/sec 16.4
;,$g ;,4; l 0 2 y,59
- 13. a.verage Mass velocity. 10 lb , fir-ft 1
il Values sed for trerea: tydracli: core a alysis (2' Tre W-1 Octrelatto- is used for analysis of some at:: *ents 1 :-;wi-- fe pr es s ar t rat ior :f ::e stem: 1; s t er , , fSee .a:;e 15.0-2, stee: In
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4 - - + - .- .. . , . i n TABLE 4.1-1 (Cont ' d) s BTRON 4KD BRA *Ds800D BYaDN AND BRAIDt000D UNITS 1 and 2 DMITS 1 and 2 WCAP-9503 l OPTIMi1ED FUEL REFERENCE DESIGg (Inni PULASITIO FUEL) EERMAL AND IfYDRAULIC DESIGN PARAMETERS COO
- ANT TEMPERATURE, k 556.9 559.2 561.6 l
- 14. Naa:nal Inlet 61.1 58.4 58.5
- 15. Average Rise in Vessel 63.6 60.7 61.8
- 16. Ave: age Rise in Core 592.4 - 591.1 594.2
- 17. Average in Core
- 18. Ave: age in vessel 55 .4 588.4 592.3 H EAT TR AN SFER a
57,500 57,500
- 19. A ive Beat Tr ansf er , Sa:! ace Area, it 2 $9,.30 197,200 197,200
- 20. Average Seat Flan, Btu /hr-f t 2 181,500 3
- 21. Mas;m an Bent Flus f or No;aal Operation, "
0 9 44 ,333 g 457,500 457,500
- Bt T.r-ft*
- 22. Ave: nge Line ar Power , aw f t '5.44 5.44 5.44 h 2 3. Pe as
- tnear Power f or NO;3 31 Oper ation, 10.f 12.6 12.6
<W f t*)
- 24. P e s * *.i ne s: Power T.msalt:n; fros 3eerpower Tr ansients 'Cperator Errc:s tassasing a 15.0 16.0 18.0 ca s ;i an cr e pow er of 1156 , <W/ft'**3 25 Fest ".i i e s: Power for 7: eve". tion of > 16 * ** 318.C >16*-
Cente:Line Melt, kW/ft(***)
- 7t;s _ia:: 2 e as s>::atet .:-' the val;e of F; = 2.32
** S e+ 3 ;:s er:t an 4. 3. 2. 2.6
*** See 5;bse r: s se. 4.4.2.11.4.
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_ - . - - m._. _ . . _ . _ . _ . _ . . . . _ . e 1 3 4 TAB LE 4.1-1 (Cont'd) 1 l BYRON AND BRAIDesOOD BYROff AND BRAIDWOOD UNITS 1 and 2 UNITS I and 2 WCAP-9500 THERMAL _ AND tTYDRAULIC DESIGN PARAMETERS (thW PARASITICfDEL) (OPTIag! ZED PUEL) g!;REEE DESIGN
- 26. Power Density, kW per Liter of CoreI *I 104.5 104.5 104.5 2'. Specific Power, kW per kg Uranium 38.4 41.9 41.9 DEL CENTRAL TEMPERATURE
- 28. Peak at Peat Linear Power f or Prevention of Centerline Melt, O r 4700 4703 4700 li. PressJre DropI **I Aer oss Core, psi 26.9:2.7E ***' 26.3* .6 25, -2.6 f I ***E 45, A::sss Vessel, Inc1J3ing Nozzle psi 47.4+4.7 46.4+4.6 4.6 ,
i ?L CSRE MECHANICAL DE31GN PARAMETERS > 1
- x
.*f
- RCC Ca71ess RCC Ca11ess 3CC Ca71ess
! . Design 17 a 17 17 s 17 s
. 17 193 193 193 3'.. N.mber of Fuel Lasem: lies 4
264 264 264 ; II. 3C2 R de per Assembly j '
- 13. F oi Pa t:h, t r. . .496 0.496 0.496
' 34. Overall Dimensions, in. S.426 s 8.426 8.426's 8.426 .4:6 s 8.4:6
- 35. Pael neight tas UO y.', lo ::.,'3)
- 234,236 :;4, 36
! ! .913 43,3't 43,3'6 it. C'. 33 Weight, It I
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s TABLE 4.4-1 TRERMAL AND BYDRAULIC COMPARISON TABLE BTRON AND BRAID 0DOD WCAP 9500 BTROM AND BRANnD e UNITS 1 AND 2 kEFERENCE UNITS 1 AND I l DESION PARAMETERS . _ IDW PARASITIC FUEL DESIGBI (OPT 2 MITED FDE i_ Reactor Core Beat Output (1004), MWt 3411 3411 3411 Reactor Core Beat Output, 10' Btvhr 11641.7 11641.7 11641.7 Beat Ger.er ated a n Fuel, 4 97.4 97.4 97.4 Sys tem. Pressur e, Ncusinal, psia II 2250 2280 2280 Sys tem Pr ess ar e, Minimm S teady-S tate, psia II 2220 2250 225C M i r. i z Jb DNBR at Nominal Conditaons II Typical Flow Channel 2.09 2.4C 2.4' Thantle (Cold Wall) Flow Channel 1.74 :. ( 2.32
. M a r::E.us DNBR f or Design Transients III ' E C
e g Typs cal, Flow Channel 31.30 . 49 31.49 4 u Thiatie Plow Channel >l.30 _. 7 >l.47 g CNS C:rrelat&on *F* (W*I wit: WR3-1 WRB-1 Modified Sparer factor) C X : ANT P1hw 6 T :: a '. TDersal Plow Rate, 10 lb,/hr .38.6 - 144.9 E!!er-1ve riow Eate f or Heat '. ): . 4 ".23." 6 Transfer, 10 lb,/hr Effective Plo,w Area for Beat $1.1 54.1 54.1 T!a*Sfer, ft' Aver age Velocity Alorig Fael 16.4 i 16.2 Rods, ft/sec 6 2 A6e: age Mass Ve;ocity, 10 g3 yjg,_gg 2.15 _ ti 2.56
a TABLE 4.4-1 (Continued T_H_EAM. AL AND __H.YDRAULIC COMPARI_SDN TA.EL;.E EYRON AND BRAIIW30D M2AP 9500 BYRON AND BRAIDWOOD UN!TS 1 AND 2 RffERENCE UNITS 1 AND 2 l EEE @ .FA M S g R,@ LM fA. RAS ((((,F"7EL _ DES : GN JPTIMI 2 F.DJELL
, C__W ANT TEMPERATURE Swanal Inlet, F 556.9 tei.9 559.2 !
A.erage Rise in Vessel, OF 61.1 56 . - 58.4 A.e: age Rise in Core, OF 63.6 -
'60.'
Asera;e in Core, F 590.4 H. 54?.1 Average in Vessel, F 587.4 32., S96.4 E e
- lit.AT "CANSFER S
?. .
* # 57,500 Ac t . . e fic a t Tr ansf er , Sar f a:e trea, ft 55,'C0 :'.500 A.er n,e Heat F 1.x , htu't.r-f: 169,6CJ .P,'00 19 7, .'0 C M a n . i .: .9 e a t F1.m 'o: Not m a; 44 ,300 4!~ 5.s 0 457,500
Le r s . or. ht; 'tr-ft A '. r . . reor is=er, ab/ft .44 .44 t.44 rsi. e a: Power f ar N ar e t. 1;.e 4.f 12.6
- e
- st.n,, nW t: 'l ieae .. ear Fwer Res6 ting ' ' .it n e : pewer le.. 16.0
.sa t a . ' Ope : a t c,r Errors ass a. . a ass.s n I
. . < : p . t- r of 118ts, t h .' f t
- _. var P ow e r for Preve :.2r s' Tenter!:ne a.9.C .S.: >18.C I'**'
... ..It
; -e: : .u t : == pe: 1 ster * : ore .;4.. .4.* 104.-
S;e . F . w .p r , ah per kg C1.J' 36.4 41.L 41.9
_ . . . . . . 4_.. , . - _ . , _ .. _ _ _ . .,.. . T e J TABLE 4.4-1 (Continued) s THERMAL AND HYDRAULIC COstPARISON TAbl.E ; 1 STROM AND BRAIDefDOD WCAP 9500 BYitOh AND 3RAIDMDOO ' ONITS 1 AED 2 REFERENCE UNITS 1 AX3 2- l DEF IGlu FU AMETERS thef PARASITIC FUEL DESIGN JCPTIMIIE:: FUE:. -- 4 1
' FUE . CEWRAL TEMPFJtAWRE i
4 4 Peak at Peak Linear Power for Prevention 4700 4700- 4700' i of Centerline Melt, OF Pressure Orop I '*I
.Across Core, psi 26.9 f, 2.7* " .7 + 2.6 26.3 + 2.4 l
- Across Vessel, 1ecluding' nozzle. psi 47.4 + 4.7'** 45.7 + 4.6 46.4 + 4.f C
N ot 4 i im i j m i i.40 t 1 4 3 1 Tr.s lanat is assoc:sted watt. the va!Je of Fg a ;.32
- ** See 5 2bsec tion 4. 3. 2.2.6.
*** See S .os ec t ia . 4.4.2.11.6.
+ bemed on co;d C aensions and 95% of tr.ec:e t ic ; density ta; i
+- 54:e3 w. vest estasate r e ac; tor fle. rate as d: gras se-d a n Se -- :- 5 1
+" P:essare Drops .pd a t e 3 :,a s ed on rest.Its from Eeference 5.
- ta 1:ese n abers are not di r ectly compar atie f or each p;a.c at.ig- d.+ : .,
t?e iv orporatts. of a di!!erent sters.a; des t;n pr wed are at t M5 or re;at ion i, 1- ite ;:efert core. . 4 2 V41.e 2 sed for :tersal hydra.alic core analysas. 1 i r 4 4 L f
J 630 620 -
=.
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/
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, /
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- = T. -
557.00 F [ 550
- I I I I
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. .p.
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B/B-FSAR 9 operating plants have indicated that the actual ficw has been well above the ficw specified for the thermal design or the plant. By applying the design procedure described in the following, it is possible to specify the expected operating flow with reasonable accuracy. Three reactor coolant flow rates are identified for the various plant design considerations. The definitions of these flcus a re presented in the following paragraphs. Best Estimate Flow The best estimate flow is the most likely value for the actual plant operating condition. This flow is based on the t+st estimate of the reactor ves,sel, steam generator anc piping flow resistance, and on the best estimate of the reactor coola:.t pump ., head-flow capacity, with no uncertainties assigned to eit1.>r th_ system flow resistance or the pump head. Syst em pre;- .;;e droir, based on best estimate flow, are presented u. r uel- .
-2 Although the best estimate flow is the mort lu ul; v; u t; :"
expected in operation, more conservative flew ra _ t :e a; plied in the thermal and mechanical designs. Thermal Design Flow Thermal design flow is the basis for the reactor core chermal perf ormance, the steam generator thermal performance, and tne nominal plant parametera used throughcut the design. To provile the required margin, the thermal design flow accounts f or th : uncertainties in reactor vessel, steam generator and pirir.J tlos resistances, reactor coolant pump head, anc the methoas use; to measure flow rate. The thermal design flow is approximateli 5.9A l less than the best estimate flow. The thermal cesign flow is confirmed when the plant is placed in operation. Iabulationc of important design and performance characteristics of the reactor coolant systems, as provided in Table 5.1- 1, are based on ti.e thermal design flow. Mechanical Design Flow Mechanical design flow is the conservatively high flow used in the mechanical design of the reactcr vessel internals and fuel assemblies. To ensure that a conservatively high flow is specified, the mechanical design ficw is cased on a reduced system resistance and on increased pump head capability. The mechanical design flow is approximately 3.7% greater than the l best estimate flow. Maximum pump overspeed results in a peak reactor coolant flow of 120% of the mechanical design ilow. This overspeed condition, which is coincident with a turbine generator overspeed of 20%, is only applicable if, when a turnine trip would be acttated, the turbine governor f ails and the turcine is tripped by the mechanical overspeed trip device. 5.1-4
. B/B-PSAR ' TABLE 5.l'-l SYSTEM DESIGN AND OPElGTING PAI:AMET!:RM Plant design life, years 40 Nominal operating pressure, psia 2250 Total system volume including 12,071
, 3 pressurizer and surge line, ft System liquid volume, including 11,695 pressurizer water at maximum 3 guaranteed power, ft Pressurizer spray rate, gpm 900 Pressurizer heater capacity, kW 1800 Pressurizer relief tank volume, ft 1900 SYSTEM TilERMAL AND IIYDRAULIC DATA (Based on Thermal Design . Plow) t 4 PUMPS 3 PUMPS ** I RUNNING RUNNING NSSS power, MWt 3425 2569 Reactor power, MWt , 3411 2560 4 Thermal design flows, gpm* Active loop 94,400 98,000 l Idle loop -- 0 Reactor 377,600 294,000 Total reactor' flow, 10 lb/hr 140.3 110,5 , Temperatures, T Reactor vessel outlet 618.4 612.2 Reactor vessel inlet 558.4 552.3 5.1-8
B/B-PSAR TABLE 5.1-1 (Cont'd) Steam generator outlet 558.1 552.1 Steam generator steam , 543.3 538.0 Feedwater 440 408.0 Steam pressure, psia 990 947 6 Total' steam flow, 10 lb/hr 15.13 10.84 Best estimate flows, gpm* Active loop 100,300 105,300 l Idle loop -- 0 Reactor 401,200 315,900 l Mechanical design flows, gpm* Active loop 104,000 109,500 l* Idle loop -- 0 Reactor 416,000 328,500 l SYSTEM PRESSURE DROPS (Dased on Pour-Loop Best 1: stimate Flow) Reactor vessel AP 44.7 Steam generator AP, psi 38.3 Hot let piping op, Psi 2.3 Pump suction piping AP, psi 3.2 Cold leg piping AP, psi 2.3 Pump head, feet *290 l
*At pump discharge. ** Calculated assuming all feedwater enters steam generatoru through the main feedwater nozzle.
5.1-9
fl0TES TO FIGUF.L 5.l-2 MODE A STEADY-STATE Ft;tt POL:ER OPERAi!0'! PRESSURE TEMPERATURE FLOW LOCAT! Oil FLUID PSIG 'F GDM II) 1R/UR I2) VC'_ ' *
'l R.C. 2235.0 617.9 112,259 37.35 -
,2 2233.1 617.9 112,159 37.3? -
3 2195.9 556.7 100,300 37.33 - 4 2192.4 556.7 '7 100,494( 6 40 25 228C.1 556.9 100,400- 37.'0 -
'6 2283.2 556.9 100,300 37.36 -
7(3) .
-2234.1 617.9 100 0.0333 -
S I4) " 2285.1 556.9 100 0.0 371 - 9 2194.2 587.0 193 0.0701 - 10-18 SEE LOOP #1 SPECIFICtJICh5 - 19-27 SEE LC'OP #1 SFECIFICAT:C Z - 28-36
" SEE LOOP #1 SPECIFICATIC:;S -
\
~ 17 2285.1 556.9 1.0 0.0004 38 2285.1 556.9 1.0 0.000:
39 2235.0 556.9 2.0 0.0008 40 STEAM 2235.0 652.7 - -
. 720 41 R.C. 2235.0 652.7 - - 10E0 42 2235.0 652.7 2.5 0.0008 -
43 2235.0 652.7 2.5 0.0008 - 44 STEAM 2235.0 652.7 0 0 - 45 R.C. 2235.0 <652.7 0 0 Mitil!!!ZE 46 N 2 3.0 120 0 0 - 47 R.C. 2235.0 <652.7 0 0 M!ti!"IZE ( 48 N 2 3.0 120 0 0 - 49 3.0 120 0 0 - 50 3.0 120 - - 450 51 PRT 3.0 120 - - 1350 WATER (1) At the conditions specified.
, (2) X 10 6 (3) Location point refers to the three 1" connections on the hot leg.
(4) location point refers to the 2" connection on the cold le9 3 M si' recc AT (MC.ul.ii',(2.s BYRON /BR AIDWOOD ST ATIONS Of besT (1sh A4 d e fh FIN AL S AFETY AN ALYSIS REPORT is 100, '.0 0 cy m . (f.clu es on 1ks hb w:ald eny FIGURE 5.1-2 harain,ma.lly a f Q cA e ,! REACTOR C00LAMT SYSTE"
- PROCESS FLOW OIAGRA:t lY (cc,afc4fJIUl .)
(SHEET 2 of 2) Ided 1ed 6p'**3 t 2
s B/B-FSAR 5.4.2.5.3 Mechanical and Flow-Induced Vibration Under Normal Operation In the design of Westinghouse steam generators, the potenti.al for tube wall degradation attributabic to mechanical o'r ' flow-induced excitation has been thoroughly evaluated. The e 21-untion included detailed analyses of the tune suppor t systeam for various mechanisms of tube vibration. The primary cause of tube vibration in heat exchangers in hydrodynamic excitation-due to secondary fluia riow on th" outside of the tubes. In the range of normal ateam generator operating conditions, the' effects of primary-fluid flew incide the tubes and mechanically induced tube vibchtion are considered to be negligible. To evaluate flow induced tube vibration in the preheater region of the tube bundle, Westinghouse undertook dn extensive program employing data from operating plants, full and partial scale model tests, and analytical tube vibration models. Operating plant data consisted of tube wear data from pulled tube eval-uations and eddy current tests and tube motion data from accelerometers installed inside celected tubes. Model testing generated tube wear data, flow velocity distributions, tube motion parameters, and flow-induced tube vibration forcing functions. The tube vibration analyses applied the forcing functions to produce tube motion data. The results of this evaluation were consist.ent with the early operating experience of preheat steam generators. On the basis of an extensive model test and analysis program, Westinghouse designed, verified, and implemented a modification to the steam generator to reduce tube vibratory response to preheater inlet flow excitation. Additionally, the magnitude of the flow forcing function was reduced through implementation of a preheater flow bypass arrangement in the feedwater system. The verification of the performance of'the modifications in reducing tube excitation and response was.done with input from a full-scale test under simulated conservative' flow and tube support conditions. 's
, s Fatigue of the tubes in ~thIe preheat'er region which are subject ~~
to flow-induced excitation is not a concern since the malimum ' resultant stresses in the tube are beloF the endurance limit of the material. s, l For areas of the tube bundle other than the preheater, parallel flow analyses were performed to determine the vibratory deflections. These analyses indicate that the flow v'locities e are sufficiently low such that they result in negligible fatigue and vibratory amplitudes. The support _ system, therefore, is deemed adequate with regard to parallel flow excitation.
- 5. t-15 N-t -
~
B/B-FSAR To evaluate crossflow at the exit of the downcomer flow to the tube bundle and at the top of the bundle in the U-bend area, Westinghouse performed an experimental research program of crossflow in tube arrays with the specific parameters of the steam generator. Air and water model tests . ore empl:/ed. The results of this research indicate that these regions of the bundle are not subject to the vortex chedding mechanism of tube excitation. Vortex enedding was found not to be a significant mechanism in these two regions for the following reasons:
- a. Flow turbulence in the downcomer and tube bundle inlet region inhibit the formation of von Karman vorticies.
b'. Both axial and crossflow velocity components ! exist on the tubes. The axial flow component I disrupts the Von Karman vortices. This research program was also the basis for evaluation of the fluid-elastic mechanism due to cross flow at the tubesheet. The evaluation showed the adequacy of the tube support arrangement. Flow turbulence can result in some tube excitation in these regions. This excitation is of little concern, however, since:
- a. Maximum stresses in the tubes are at least an order of magnitude below the fatigue endurance limit of the tube material, and
- b. Tube support arrangements preclude significant vibratory motion.
In summary, tube vibration has been thoroughly evaluated. Mechanical and primary flow excitation are considered negligible. Secondary flow excitation has been evaluated. From this evaluation, it is concluded that if tube vibration does occur, the magnitude will be limited. Tube fatigue due to the vibration is judged to be negligible. Any tube wear resulting from the tube vibration would be limited and would progress slowly. This allows use of a periodic tube inservice inspection program for detection and followup of any tube wear. This inservice inspection program, in conjunction with tube plugging criteria, provides for safe operation of the steam generators. 5.4-16
B/B-FSAR ~
- 5. 4. 2. 5. 4 All'owable Tube Wall Thinning Under Accident Conditions An evaluation is performed to determine the extent of tube wall thinning that can be tolerated under accident conditions. Under such a postulated design-basis accident, vibration is of short enough duration that enere is no endurance problem. The results of a study made on "D series" (.75 inch nominal diameter .043 inch nominal thickness) tubes under accident loading are discussed in WCAP-7832 (Reference 3) and show that a minimum wall thickness of .026 inches would have a maximum f aulted condition stress (i.e., due to combined LOCA and safe shutdotin earth-{uake loads) that is less than the allowable limit. This thicknr:sa is
.010 inches less than the minimum steam generator tube wall thickness .039 reduced to .036 inches by the assumed general corrosion and erosica loss of .0033 inches. O The corrosion rate is based c. a conservative weight loss rate for Inconel tubing in flowing 6500 E primary side reactor coolant fluid. The weight loss, when equated to a thinning rate and projected over a 40-year plant life with appropriate reduction after initial hours, is equivalent to .083 mils thinning. The assumed corrosion rate of 3 mils leaves a conservative 2.917 mils for general corrosion thinning on the secondary side.
The steam generator tubes, existing originally at their minimum wall thickness and reduced by a very conservative general corrosion loss, still provide quite an adequate safety margin. Thus, it can be concluded that the ability of the steam generator tubes to withstand accident loadings is not af fected by a lifetime of general corrosion losses. 5.4-16a !
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B/B-FSAR and under startup and light load conditionc when the prehect": section is bypassed. The water hammer preventive features are more fully decerihed in the section that follows. As shown on Figure 10.1-1, the valves and the piping downat. m:r thereof are Safety Category I, Quality Jroup B. The valves and their downstream sections of Category I main feedwater and temporing piping are located in the same Category I valve rooms which N;use the main steamline isolation valves described in Section 10.3. 10.4.7.3 Water Hammer Prevention Features Several water hammer prevention features have been designed into the feedwater system. These features are provided te nini-i a : the possibility of various water hammer phenomena in the 't- 1:1 l generator preheater, steam generator main feedwater irl,t pi, p1 + and the steam generator upper nozzle feedwater piping. T: ;tlTwing , discussion is typical for each of the four steam genu ators and ; their associated feedwater piping. j s 10.4.7.3.1 Start-Up, Low Load Conditions !
- a. Under start-up and low load conditions when NSSS rated flow is less than 15% and temperatures are less than 250' F, feedwater will only be admitted to the upper !
nozzle of the steam generator by the use of flow through i the feedwater bypass tempering line and/or flow through l the feedwater preheater bypass line via the feedwater l bypass control valve and feedwater preheater bypass valve. I The 6-inch diameter upper nozzle is located on the upper i shell of the steam generator, below the normal, full power water level. Level control in the steam generator is provided by the feedwater bypass control valve at these conditions.
- b. Surface mounted resistance temperature detectors (RTD) j are provided on each of the feedwater pipes, leading j to and very near the steam generator's upper nozzle to detect during start-up and low load conditions as well l i
as other operating conditions, possible back leakage l of steam from tue steam generator into the feedwater piping. These RTD's are monitored by the plant proccas computer and alarmed in the main control room so that actions can be taken to initiate feedwater flow to the upper nozzle before potential feedwater hammer conditions may develop. 10.4.7.3.2 Increasing Load
- a. As load increases about 15% of NSSS rated flow and feedwater .
temperatures rise above 250a F, forward feedwater flushina of the main feedwater piping may be initiated by opening l 10.4-11
B/B-FSAR the feedwater isolation bypass valve. A small contro11 d l flow through the 3-inch f eedwater isolation bypass line j is provided to flush the main feedwater piping between I the isolation valve and the steam generator.
- b. Three sets of three RTD's are provided on the main feed- l '
water piping upstream and acwnstream of the f eetnter j isolation valve and near the steam generator f eedur.t s t . nozzle to detect when the feedwater flushing temp 3rature rises above 255 F. Two ou t of thr ee logic is providad for each set of three RTD's and all three must be sat!:fied : I to meet the forward flushing temperature cequirement .
- c. If flow in the 3-inch feedwater isolation valve byp*1s ,
line (forward flushing flow) remains above a prenet ainimum and below a preset maximum and the flushing tamperatu >s remain satisfied, a timed period occurs af ter wnich a :
. permissive signal is provided to automati: ally Span the '
.feedwater isolation valves. Automatic oneninq fi f"71-water isolation valve can be blocked by plaaing i t r. asa. col switch in the main control room in the cic:ca position. i This automatic permissive to open occurs a'ter a timed !
period to allow epproximately two volumes et water to ! be purged from the piping between the f ee tiva te r iso i nt ion ' valve and the steam generator main feedwater nozn12. 1 Feedwater flow at the main f eedwater flow-element maat i also be above a precet minimum in order for the feedwater j isolation valve to open. l
- d. Af ter the feedwater isolation valve has opened, the feed-I water isolation bypass valve will be manually closed.
l
- e. Prior to opening of the feedwater isolation valvo, transfer f rom the f eedwater bypass control valve to the f eet. s ter control valve will occur in order to provide steam g2nerator level control at the higher feedwater flow conditions.
- f. If flow to the steam generators remains continuous during i a load transient and above a minimum flow rate, feedwater !
will not be terminated to the main feedwater nozzle even if temperature of the feedwater has dropped below 250* F. l Interruption or a reduction in flow below the minimum i rate however, will cause the feedwater preheater certion ; of the steam generator to be bypassed.
- g. Steam generator low level trips are provided to close all of the feedwater isolation valves, feedwater isolation bypass valves and feedwater preheater bypass valves.
Steam generator low pressure trips are provided to close all of the feedwater isolation valves, feedwater isolation l bypass valves, feedwater preheater bypass valves and i the feedwater bypass tempering valves. 10.4-12
. D/B-FSAR 10.4.7.3.3 Split Feedwater Flow l l
- a. Prior to opening of the feedwater icolation valve, the ,
majority of feedwater flow at the lower power level ic j introduced to the upper nozzle of the steam gene ator ; by the preheater bypass pipe. ,
- b. At higher power levels af ter the feedwater icolation valve has opened, only a small portion of the feedwater ,
flow bypasses the preheater , with the hypasa par tien contributing to approximately 101 of full foot >ater flow ; t at 100% power. This split feedwater finw accango~.n: ' provides an approximate 901 of full Elcw limit to th7 main feedwater nozzle at higher power lev 212 in or br to minimize the potential for tubing vibration in tSe steam generator. The feedwater f ics. ratn *c the e - generator nozzle is monitored and alarmed, if tio~ r i : 2:- above approximately 901, in order for act'c' . ta Sc . 9 =r to reduce flow. I
- c. The preh6ater bypass valve remains open th r oug hou t the '
start-up and low load conditions, as well as up to and including full power operation. 10.4.7.3.4 Other Upoer Mozzle Feedwater Line Uses Ina.smuch as there . is water flowing to the upper nozzle of the i steam generator during normal operation, and it is the required location for introducing cold fluid into the steam generator , ! auxiliary feeedwater and chemical feed are connected to the upper nozzle feedwater lines rather than to the main feedwater linen. 1 The chemical feed lines are used to add chemicals directly to i the steam generators under low load conditions prior to wet layup. The chemical feed and auxiliary feedwater lines are Safety Category I, Quality Group B out to, and including their isolation valves. 10.4.7.4 Safety Evaluation The condensate and feedwater systems are not safety-related except i as described in Subsection 10.4.7.1.1. If it is necessary to remove a component such as a feedwater heater, pump, or control valve from service, continued operation of the system is possible by use of the multistream arrangement and the provicions for removing f rom service and bypassing equipment and uactions of the system. An abnormal operational transient analysis of the loss of a feed-water heater string is included in Subsection 15.1.1. 10.4-12a
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errors in the determination of the steady-state power level are made as described in Section 15.0.3.2. The thermal power values used for each transient analyzed are given in Table 15.0-2. In all cases where the 3579 megawatt thermal (MWt) rating is used in an analysis, the resulting transients and consequences are conservative compared to using the 3425 MWt rating. The values of other pertinent plant parameters utilized in the accident analyses are given in Tables 15.0-3 and 15.0-4. 15.0.3.2 Initial Conditions For most accidents which are DNB limited, nominal values of initial conditions are assumed-(including an appropriate tem-perature margin to compensate for steam generator tube fouling). The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the limit DNBR, as described in WCAP-8567 (Reference 10). This procedure is known as the " Improved Thermal Design Procedure," and is discussed more fully in Section 4.4. For accidents which are not DNB limited, or in which the Improved Thermal Design Procedure is not employed the initial conditions are obtained by adding the maximum steady state errors to rated values. The following conservative steady state errors were assumed in the analysis:
- a. Core Power 12% allowance for calori-metric error
- b. Average Reactor i 4.90F allowance for l Coolant System controller deadband and temperature measurement error and steam generator fouling penalty
- c. Pressurizer i 30 pounds per square inch pressure (psi) allowance for steady state fluctuations and mea-surement error Table 15.0-2 summarizes initial conditions and computer codes used in the accident analysis, and shows which accidents employed a DNB analysis using the improved thermal design pro-cedure.
15.0.3.3 Power Distribution The transient response of the reactor system is dependent on the initial power distribution. The nuclear design of the reactor core m;nimizes adverse power distribution through the placement of ccatrol rods and operating instructions. Power distribution may be characterized by the radial factor (FAH) and the total peaking factor (FQ). The peaking factor limits are given in the technical specifications. 15.n-n
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