ML20217A511
| ML20217A511 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/24/1998 |
| From: | Byrd M, Richerson M TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20217A423 | List: |
| References | |
| GL-96-06, GL-96-6, MD-Q-0999-92013, MD-Q-0999-920136-R04, MD-Q-999-92013, MD-Q-999-920136-R4, NUDOCS 9804220274 | |
| Download: ML20217A511 (10) | |
Text
. .
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 REQUEST FOR ADDITIONAL INFORMATION REGARDING (RAI) RESPONSE TO NRC GENERIC LETTER (GL) 96-06, ASSURANCE OF EQUIPMENT OPERABILITY !
AND CONTAINMENT INTEGRITY DURING DESIGN BASIS ACCIDENTS i
TVA Calculation MD-O-0999-920136, Revision 4 l
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9804220274 98041S PDR ADOCK0500g9 P
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, TVAN CAL.CULATION COVERSHEET
--- Title Post Accident Primary Containment Pene: ration 1%21ated Piping Over Page ic of 14 Pressure Analysis Plant BFN Unit 1.2.3 Prepanng Organization Key Nouns (For RIMS)
GE 3458. picing, pressure, entmt. penetration Branch / Project identifiers Each time tnese calculations are issued, preparer must ensure that the onginal(RO)
RIMS accession number is filled in.
MD-Q0999-920136 i
Rev (for RIMS use) RIMS Accession Number Appleable Design Document (s)
N/A RO R22 92 0605113 R4 8^
R14 980324 10T
- 'g UNID System (s) R5 Section(s): 5.2 vanous R6 Revision 0 R4 R R Quainty Related?
Yes No Design Change Document No.
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Not Applicable Sareey reiated?
(or indicate Not Applicable) Yes No g g Prepared MWMark Renerson tw calculahans contain Yes No VB unweed assumpton(s) tnat rnust
% m w.n O 2 Checked these caicuiatens contain specia' LS
- 8#0 """*** Yes No ma-w" *"d' '"""'"8 iy O 2 Reviewed AG These casculatone contain a Yes No d"'on output suscrmnt?
Approved O 2 Calculation Rowson:
HC Enta Calculaten Date
~
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Q Selected pages @
6-5-92 BJ p['fy# Not Apphest*
O ,
Statement of Problem:
/
Evaluate the primary containment penetration thermal over pressure issue l per generic letter 96-06.
Abstract ORIGINAL:
See continuation page.
This calculation is applicable to power operation at 3293 MWt and power uprate at 3458 MWt.
O Microfilm and retum calculation to Calculation Library. Address: O Microrism and destroy.
O Microfilm aro retum calcuistion to:
TVA 40532 (08-97] NEDP-2-1 [08-05 97)
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, TVAN CALCULATION COVERSHEET Title Post Accident Primary Containment Penetration Isetated Piping Over Page ic of 14 Pressure Analysis Prepanng Organization Plant BFN Unit 1.2.3 Key Nouns (For RIMS)
GE 3458. piping, pressure, entmt. penetration Branch / Project identifiers Each tirne these calculacons are issued. preparer must ensure that the ongenal (RO)
RIMS accesse number is Riled m.
MD-Q0999-920136 Rev (for RIMS use) RIMS Accession Number Apphcaole Design Document (s)
N/A R0 R22 92 0605113 R4 SA a ect
, UNID System (s) R5 Section(s): 5.2 various R6 Revirston 0 R4 R R oueury Romeo? Yes No Desgn Change Document No. E O Not Apphcable Serety reisted?
(orindicate Not Applicable) Yes No 2 O Prepared vs g These ceicumione contan y, g Checked
. ., 2 .p~ ~ ~ O e These concumens coniaan spec
- LS 'Yes No
, recurements and/orinting RM O S AG These calcuestione contam e Yes No design output attachment?
O 3 APP'0V*d Calculaban Revision:
HC Entre calcuistion O
82 8 Statement of Problem:
Evaluate the primary containment penetration thermal over pressure issue per generic letter 96-06.
1 Abstract
)
See continuation page.
j This calculatien is applicable to power operation at 3293 MWt and power uprate at 3458 MWt. l O Micrortim and retum calculation to Calculation Ubrary. Address: C Mscronim and destroy O Microfilm and retum calcuistion to:
TVA 40532 [06-97] NEDP 21,^44 * '
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s TENNESSEE VALLEY AUTHORITY Sheet i-a
SUBJECT:
CALCULATION MD-OO999-920136 REV. 2 PROJECT: BFNP ,
" POST ACCIDENT PRIMARY CONTAINMENT PENETRATION ISOLATED PIPING OVERPRESSURE ANALYSIS" REVISION 2: PREPARED <rW DATE 5' 97 CHECKED 'M DATE T/2//V7 COVERSHEET CONTIUATION:
ABSTRACT Generic Letter 96-06 identified a concern that thermally induced over-pressurization ofisolated water filled piping sections in containment could jeopardize the ability of accident mitigating systems to "
perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. This calculation evaluated the mechanical piping systems that penetrate the primary contamment to determine if any are susceptible to the postulated thermal expansion of trapped water such that over-pressurization of the piping pressure boundary could occur.
The primary containment penetrations were reviewed to determine the piping penetrations which could be susceptible to the thermal pressurization. Based on the established elimination criteria in Section 5.0, the majority of primary containment penetrations were dispositioned as not being susceptible to the thermal pressurization. Four primary containment penetrations and process lines were identified as being susceptible to thermal pressurization; 1) demineralized water system, 2) drywell floor drain sump discharge, 3) drywell equipment drain sump discharge, and 4) reactor water sampling system.
These lines were evaluated to determine if the current configurations are acceptable. The demineralized water system is acceptable as-is provided water is drained from the piping prior to I power operation each cycle. The drywell floor and equipment drain sump discharge lines are considered acceptable based on leakage of nonsafety-related valves inside containment. The thermally induce pressurization of the reactor samplin'g system is calculated to reach approximately 2,546 psig prior to the lifting of the inboard globe isolation valve disc offits seat and relieving the pressure the reactor vessel.
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. TENNESSEE VALLEY AUTHORITY
SUBJECT:
Sheet 6 of Lt CALCUITATION MD-OO999-920136 REV. 2
. PROJECT: BFNP -
" POST ACCIDENT PRIMARY CONTAINMENT PENETRATION ISOLATED PIPING OVERPRESSUl@ ANALYSIS" REVISION 2: PREPARED SDP DATE 5 97 CHECKED h- DATE f//r/47 4.0 INPUT DATA Parameter Input Data Reference Reactor Water Sampling System:
l Pipe Material Seamless A312 or A376, 3.19,3.20 Grade TP304 or TP316 Penetration Process Pipe Material A312,TP304 3.21,3.22 1
Process Line Size, Schedule 3/4 inch, Schedule 80 3.19 Pipe Outer Diameter (D.) 1.05 inches 3.24 PipeNominal Thickness 0.154 inches 3.24 l Manufacturer's Tolerance 12.5 % 3.23 Drywell Maximum Temperature 322*F 3.16 1
Maximum Allowable Stress (SE) 15,418 psi 3.25 at 322
- F ,
Isolation Valve Spring Pre-load 800 lbf Attachment A Isolation Valve Disc Diameter 3/4 inch Attachment A Isolation Valve Stem Diameter 1/2 inch Attachment A Sampling System Design Pressure 1326 psig 3.17 Dem'Idifra :
x sectional area of 1 me 0.006 ft2 3.24 water density at 50*F lb,,,/ft' 3.28 322*F- 56.50lb,,,/ 3.28 length ofisolated section in drywell 400 feet (estimate)
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%O NM 3,1 i
TENNESSEE VALLEY AUTHORITY Sheet _11 of _}i,
SUBJECT:
CALCUL% TION MD-00999-920136 REV. 2 PROJECT: BFNP -
~ " POST ACCIDENT PRIMARY CONTAINMENT PENETRATION ISOLATED PIPING OVERPRESSURfANALYSIS" REVISION 2: PREPARED TOP DATE f.21-9 7 CHECKED % DATE 6/28/k7 I
5.0 EVALUATIONS (continued) l Reactor Water Sampling System I l
The reactor water sampling system takes suction off a recirculation line between the recirculation pump l discharge valve and the reactor vessel [ Reference 3.18]. The 3/4-inch sampling line has an inboard l isolation valve inside containment and an outboard isolation valve outside [ Reference 3.17]. The sampling system may be operated intermittently during normal power operation such that the isolation valves could be closed at the onset of a LOCA, or closed automatically by the primary containment isolation signal. Since the operation of the sampling system is not continuous, the water trapped between the isolated valves may not be hot at the time ofisolation, and thus subject to thermal i expansion in the event the drywell ambient temperature increased due to a LOCA.
The sampling system isolation valves are 3/4-inch,900# Class, cylinder-operated, fail closed, Y-globe i valves manufactured by Crane-Flomatics [ Reference EMPAC database and Drawing 6765 Rev. D (Attachment B)]. The globe valves have flow over the seat such that a pressure increase between the isolated valves could " lift" the inboard globe disc offits seat and relieve the pressure to the reactor vessel. Information provided by the vendor [ Reference Attachment A] detailing the disc diameter, stem diameter, and spring pre-load may be used to cdtulate the pressure required to lift the inboard isolation valve offits seat as follows [see Section 4.0 for ciesign input data]:
l Conservatively using the piping design pressure of 1326 psig, calculate above disc static force due to reactor pressure. [ Note the actual reactor pressure during a design basis LOCA is significantly below 1326 psig. However, the reactor pressure during the small break LOCA is closer to normal operating pressure.]
Force , = Pressure,m (Area of Disc - Area of Stem)
= 1326 psig (n/4) [(3/4 inch)2 - (1/2 inch)2)
= 325 lbf The thermally-induced pressure between the isolation valves required to lift the inboard globe ,
valve disc is thepressure which will exceed the force of the spring pre-load and the force of the upstream side pressure (see Section 4.0 for design input data].
Pressure w,.% = (Force, + Forced / (Area of Disc)
= (325 lbf + 800 lbf) / {(n/4)(3/4 inch)2)
= 2,546 psig
, TENNESSEE VALLEY AUTHORITY Sheet 12 ord
SUBJECT:
CALCbDATION MD-00999-920136 REV. 2 PROJECT: BFNP ,,
" " POST ACCIDENT PR} MARY CONTAINMENT PENETRATION ISOLATED PIPING OVERPRESSUREANALYSIS" REVISION 2: PREPARED 40P DATE f-7f-51 CHECKED 'fA. DATE FDA7 5.0 EVALUATIONS (continued)
The internal design pressure of the sampling system per the design basis code of record, USAS B31.1.0-1967, is calculated by the following formula (see Section 4.0 for design input data):
P = {2(SE)(tu - A)} / {Do - 2y(tu - A)}
where SE = maximum allowable stress in material due to internal pressure (psi),
ipcn. p6 e M !5,d!! pd ca 222'F for A312 or A376, Or TP304 or TP 316 S'!
tu = minimum required wall thickness (inches), includes manufactures 12.5% !
tolerance, (1 - 0.125)0.154 inches = 0.135 inches Do= piping outside diameter (inches) = 1.05 inches !
y = 0.4 (Table 104.1.2(a)2, USAS B31.1.0-1967)
A = 0.0 (Table 104.1.2(a)l, USAS B31.1.0-1967), no allowance for corrosion is added based on the Stainless Steel material type is,M 2 P = {2(4448 psi)(0.135 - 0.0)} / { l.05 - 2(0.4)(0.135 - 0.0) pg ,
= 4r149psig 4,4 o o As calculated above, the inboard globe valve will lift at approximately 2,546 psig to relieve the thermally-induced pressure to the reactor vessel. The maximum de 'gn pressure calculated for the process piping per the design basis code-of-record tiMpsig af F. The 3/4-inch 900# isolation globe valves were hydro-tested by the manufacturer to 3,250 psig [ Attachment B]. Therefore, the sampling system is susceptible to the thermal over-pressurization concern identified by GL 96-06, and is calculated to reach 2,546 psig. The thermally-induced pressure is acceptable per the mechanical design of the piping and isolation valves.
- . . _ _ _ . . . _ _ i TENNESSEE VALLEY AUTHORITY s Sheet 13 ord
SUBJECT:
CALCUL'ATION MD-00999-920136 REV. 2 PROJECT: BFNP.,
s
" POST ACCIDENT PRIMARY CONTAINMENT PENETRATION ISOLATED PIPING OVERPRESSUnfANALYSIS" REVISION 2: PREPARED 50F DATE 5-21-97 CHECKED % DATE 57Z1/47 6.0
SUMMARY
OF RESULTS The primary containment penetrations were reviewed to determine the piping penetrations which could be susceptible to thermal pressurization. Based on the established elimination criteria in Section 5.0, the majority of primary containment penetrations were dispositioned as not being susceptible to the thermal pressurization. Four primary containment penetrations and process lines were identified as being susceptible to thermal pressurization; 1) demineralized water system,2) drywell floor drain sump discharge,3) drywell equipment drain sump discharge, and 4) reactor water sampling system. These lines were evaluated to determine if the current configurations are acceptable. The demineralized water system is acceptable as-is provided water is drained from the piping prior to power operation each cycle. The drywell floor and equipment drain sump discharge lines are considered acceptable based on leakage of nonsafety-related valves inside containment. The thermally induced pressurization of the reactor sampling system is calculated to reach approximately 2,546 psig prior to the inboard globe -
isolation valve disc lifting offits seat and relieving the pressure to the reactor vessel.
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, y,, ,"* ste111te overlay (Sea t 9ing & PIvg) w.O. 94-20630 ADMIN v k svs ao.
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M S a SEATING DETAIL Ti 1 ' '~ PCADAM
..--......p PART'S LIST N0. OESCRIPfl0N OVAN. CODE WATERIAL CODE PART NO. SIZE
- 1. SPRING 1 45 CARBON STEEL DA 6138
- 2. LOCR NUT 1 17 30453 -- 50665 5/a-16
- 3. PISTON 1 55. ALVulNVWes. Og 31a12 PISTON CASTING 1 55 ALVWlHUW C8 10023
+4. 0-RINC. PISTON ROD 1 68 BVNA N -- 52014 e5. 0-RING 1 68 BVNA-N -- 52118 e6. NOT REQ'O.
h h
- 7. HEAD. CYLINDER HEAD CYLlNDER(CASTINC) 1 1 55 ALVWINVuose 55 ALVulNVW DB 31879 CB 10022
+8. 0-RING, PISTON & HEAD 2 68 BVNA-N -- 52256 e$. NOT REQ'O.
- 10. SNAP RINC 1 01 C.S.se -- 50341 h
M
- 11. 00WE 00WE 1
1 55 ALVWINVWa** 55 ALVWINUW DC 2352 OC 10021
- 12. YOKE 1 01 C.S. 09 31405 YOKE. CASTING 1 01 C.S. 00 10116
- 13. STUO. CLAND FLANCE 2 96 A -- 54610 $/16-18 X 1-5/6 LC
- 14. NVT. CLAND FLANCE 5/16-16
- 15. CLAND FLANCE
- 16. CLAND 2
1 1 to 01 A C.S... 86 31633
-- 50603 CA 10122 DA 2263 b A PK ta. f$*! $
20577-00
- 17. SQ. HD. CAP SCREW, YOKE 4 56 8 -- 50468 3/8-16 X 1" LC.
20634 00 elo. CASKET. BONNET 1 94 304 W/ ASBESTOS DA 8820/6820W Filler /er Croleil Filler M . ( bhQ"M
- 19. PACKING AS REQ'O. 47 CRANE 1871 -' 7.
1 -- 55353 8/16 SQ. .
- 20. CU10E BVSHING 1 98 STELLITE & OA 2262 d 21. BONNET 1 $6 ASTW A142-F316 08 31410
- 22. STEW. LOWER 1 to 316 W/ STELLITE 08 31403 d STEW. WELOWENT 1 la 316 W/ STELLITE 08 31408 #@o Avdhclo on h
23. 24. CACE SEAT MINC 1 1 la 316
$8 316 W/ STELLITE DA 31432 DA 31407 ANMm M dm SEAT MlNG. WELOWENT 1 98 316 W/ STELLITE DA 31406 e25. CASKET. SEAT RINC 1 94 304 W/ASSESTOS DA 31431/31431W Fifier/er Ctsfell Filler
- 26. 800Y. CA5TINC 1 $4 ASTW A351 CR-CF aWCC 10093
- 27. NUT. CYLINOER 1 17 30455 OA 2305
+28. GVf0E. PISTON R00 1 39 BRONZE DA 31478
- 29. PISTON ROD 1 19 3 SS 08 30737
- 30. STVO. INDICATOR 1 94 -- 55344 5/16-14 X 4-1/2
- 31. NUT. IN01CATOR 3 98 -- 50603 5/16-14
- 32. INDICATOR 1 01 C 3.ee DA 7276
- 33. SQ. HD. CAP SCREW. INO. 1 98 -- 50486 5/16-14 X 1-3/4 34.- PILOT VALVE e LIWIT 1 -- SEE DRAWING A8 2152 SWITCH ASSEUBLY MUD D )g DTh G.,,d ALL SVRFACES ACCESSIBLE FOR REPAIR SHALL l]E INSPECTED IN THE FINISHED COH0lT10N SY L10V10 PENETRANT WETH003 OUTLINED IN N627 0F SECTION til 0F THE ASWE 80lLER AND PRESSURE VESSEL CODE.
ALL ACCES$10LE SVRFACES SHALL BE EXAWINED BY L10V10 PENETRANT WETH005 OUTLINED IN N627 of SECTION lil CF THE ASWE BolLER AM) Leiscows WLCH ma Ws 14
- SEAT LEAKACE TEST SHALL BE PERFORWED WITH FLOW OVER THE PLVC.
TEST PRESSURE: 2200 TEST WCOIAs WATER ALLOWABLE LEAKACEs 1.5 CC/HR. i in *v m Aar u vr r 'ty set N *nt _ h , AUSTENITIC STAINLESS STEEL MATERI AL SHALL BE DECREASED AND STRIPPEG g IN ACCORDANCE WITH ASTW A360-57 PRIOR TO HEAT TREATWENT. TEWPERATURE INDICATING CRAYONS ARE NOT TO BE VSED ON AUSTENiflC STAINLESS STEELS. 11 DAG OlSP(R$10N No. 15$ WANUTACTURED BY THE ACHESON CORP. 15 THE ONI,Y LUBRICANT THAT WAY BE VSED FOR THE ASSEW8LY OF STAINLESS STEEL PARTS. gggp 3pggg pag 73 . 80LilNC WATERI AL TO BE ASTW A193-Ba. A194-6, A453 ANY CRADE. OR A461-6SS. THESE PART5 10 SE IN THE ANNEALEO CONDl110N PRIOR TO APPLYlNC THE HARO FACING ALLOT. j thC SET (Cl SVPPLY)
- ACCEPTABLE HAROFACINC WATERIALS AREe 160 AND CAN80N SPACER e) COLWON0Y 4 $ . OR 6 0 INSTALLED J-FCV-43-13 b) STELLITE 1 3. 6 HAYNES. OR STELLITE ALLOY 21 7 . ELECTRICAL SUPPLYi 120VAC 608 a 6 WElGHTl 45 lbs.
00 5 . ACTVATOR SPRING LOADED TO F All CLOSE 00 4 . ACTVAfl0N PRESSVREs 70 P.S.I.C. (WIN.) 150 P.S.I.C. (WAX.) 3 . $[RylCE: DEWINERAL12[0 WATER 2 1 WORKING PRES $VRE: 1350 P.S.I. WORKINC TEWPERATURE: 575'F. h
/242K417CTNYMAWlVt2tM CQ VAA10V3 l l KUL l l N/Al l N/Al N/A l l ph Che forms 1 INSTALLE0 NCP PACKINC SET AND CARBON $ PACER FOR 3-FCV-4313
-14 PER W.D. 94-20577-00 AND 5 O. 94-20634-00 RESPECTIVELY 00* V $ A CYLINDER OPERATOR (REF: OCN C16384): A00E0 NOTE 29 AND REYlSED UNIOS TO CONFORW TO NOTE 29 (ADWINISTRATIVE REVISION)
[(* OBE M E"SW11H SOCKET WELD ENDS Llulf SWITCHES PILOT VALVE 6 STELLITd VAR 10VS l 8-6-96 l uP T l J K l JW lRLD/Tl N/A[ ANW l 0PB lnCD/s OVERLAY 0 IS3V{D TO CREATE CCO. SUPERSEDES A/O 6765 RS01 AND ac4 [ . NONE WuDR AC. TO OLPICT AS-CONSTRUCTED STATUS PER OCN V384&7A REWOVED i oueu mi-a. as VENDOR DISCLAlWER (ADWINISTRATIVE REVISION) Robse 8/19/89
' x 3-6765 CHANGE REFl DATE l 0F T Rl CH K Rl 0 SC Nl R V W Rl' APP 0fAPPO %PPO l lsy*= SSO Q PL ALL A/D HISTORY HESEARCHED AT R000 Tm s onAsimo 13 Vuota ppn CONF 1GURATION CONTROL V V LJ J}}