3F0184-29, Forwards Minutes of 831215 Meeting Re Equipment Qualification Issue Resolution.Minutes Reflect Results of 840124 Telcon.Response to IE Info Notice 79-22 Considered in Encl 3 Methodology
ML20086L253 | |
Person / Time | |
---|---|
Site: | Crystal River |
Issue date: | 01/31/1984 |
From: | Westafer G FLORIDA POWER CORP. |
To: | Stolz J Office of Nuclear Reactor Regulation |
References | |
REF-SSINS-6835 3F0184-29, 3F184-29, IEIN-79-22, TAC-42512, NUDOCS 8402080021 | |
Download: ML20086L253 (112) | |
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January 31,1984 3F0184-29 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Crystal River Unit 3 Docket No. 50-3')2 Operating License No. DPR-72 Environmental Qualification of Electrical Equipment
Dear Sir:
Florida Power Corporation (FPC) hereby submits fifteen copies of the minutes of the December 15, 1983 meeting between FPC and NRC staff to reach final resolution of all equipment qualification issues for Crystal River Unit 3. These minutes reflect the results of the telephone discussions held January 24., 1984. In accordance with your request, the FPC response to IE Information Notice 79-22 has been considered in the methodology shown in Enclosure 3.
If you have any questions, please contact this office.
Sincerely, G. R. Westafer Manager, Nuclear Operations i
Licensing and Fuel Management AEF/feb Enclosures cc: Mr. 3. P. O'Reilly Regional Administrator, Region II Office of Inspection and Enforcement O U.S. Nuclear Regulatory Commission f 101 Marietta Street NW, Suite 2900 Atlanta, GA 30303 g
8402080021 840131 1 PDR ADOCK 05000302 P PDR GENERAL OFFICE 3201 Thirty-fourth Street South e P.O. Box 14042, St. Petersburg, Florida 33733 e 813-866-5151
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DOCKET NO. 50-302 RESOLUTION OF SAFETY EVALUATION REPORT FOR ENVIRONMENTAL QUALIFICATICN OF SAFETY RELATED ELECTRICAL EQUIPMENT FOR CRYSTAL. RIVER UNIT 3 On January 14, 1983, Florida Power Corporation (FPC) received a
. Safety Evaluation Report (SER) regarding the environmental qualification. of safety related electrical equipment at Crystal River Unit 3 (CR-3) . The SER contained a Technical Evaluation Report (TER), entitlei " Review of Licensees Resolution of Outstanding Issues from NRC Equipment Qualification Safety Evaluation Reports (F-11 and B-60) ," written by Franklin Research Center under contract to the NRC, which noted a nuniber of environmental qualification documentation deficiencies for safety i related electrical equipment at CR-3.
- On December 15, 1983 a meeting was held with members of the NRC staff to discuss FPC's proposed method of resolution for each of those deficiencies. The proposed resolutions which were discussed for each of the items in the TER are summarized in enclosure 1 to this submittal.
1
, Discussions also took place at the meeting regarding FPC's
! methodology for compliance with 10 CFR 50.49, " Environmental Qualification of Electrical Equipment Important to Safety for '
Nuclear Power Plants," which became law on February 22, 1983.
FPC responded to the law and to the Franklin Research TER in the i same document which was submitted to the NRC on May 20, 1983. In that document, Justifications for Continued Operation (JCO 's )
were resubmitted to satisfy the criteria of paragraph (i) of 10 CFR 50.49. Lubricants, cables and terminal boxes were addressed generically in that submittal. The agenda for 1
the . December 15 meeting is provided as enclosure 2 of this-
~
submittal. The documentation of the discussions which took place is provided in the'following paragraphs.
I' Paragraph (b) (2) of 10 CFR 50.49 l Florida Power Corporation does not differentiate between equipment- which is safety related and non-safety related '
equipment whose failure could prevent the proper operation of safety related equipment. If failure of a device can effect the function of safety related equipment, that device is considered safety related in the Florida Power system and is included in the >
FPC safety listing. Items defined by 10 CFR 50.49 (b) (2) are included in the safety list; however, to further ensure the accuracy of the safety listing, FPC is in the process of confirming the safety list. The methodology being used is I
. described in enclosure 3 of this submittal.
i Paragraph (b) (3) of 10 CFR 50.49 i
Florida Power Corporation has initiated a program to identify the items addressed by 10 CFR 50.49 (b) (3) and Regulatory Guide 1.97.
The schedule for completing identification of these items is l summer, 1984.
The time required to perform the necessary design and procurement activities will preclude installation of any modifications during the spring, 1985 refueling outage. FPC has made a ccommitment to make all Regulatory Guide 1.97 modifications during Refuel VI, the outage after the spring, 1985 shutdown. FPC will ask for an extension of the March 31, 1985 deadline for any additional equipment which may be identified in j the Regulatory Guide 1.97 effort.
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Justification for Continued Operation Justifications for Continued Operation were submitted by Florida Power Corporation and were reviewed by Franklin Research. In the FPC May 20, 1983 submittal, those JCO's were resubmitted to comply with the criteria of paragraph (i) of 10 CFR 50.49. One additional JCO (which had not previously been reviewed by the NRC,) was provided in the May 20, 1983 submittal.
The circumstance surrounding this JCO is discussed in note 1 to enclosure 1. The JCO itself is provided as enclosure 4. (Other items from the May 20, 1983 submittal referenced herein are included in enclosure 5).
Limitorque Valve Operators Florida Power Corporation instituted a comprehensive program to assure that Limitorque Valve Operators are qualified for the environment in which they are located. The Crystal River 3 plant was divided into zones with uniform environments. The normal and accident environments for each of these zones were established.
This information was provided to Limitorque. Limitorque l indicated to FPC which types of Limitorque operators would be l
qualified for each environment and which generic test report l applied to the various operators and zones. Limitorque has three l
generic test reports for valve operators to be installed in nuclear power plants, one for inside containment and two for
! outside containment. Limitorque maintains evidence of qualification based on the serial number of the unit.
Qualification for certain parameters can be checked by ascertaining the color of the parts of the operators. For
! example, the color of limit switches and torque switches 1
indicates their qualification. Nameplates on motors indicate the insulaticn class used on motor windings.
3
FPC has determined which test reports apply to each of the Limitorque valve operators by serial number and has performed a walkdown which encompassed each valve operator. Based on this evaluation, it has been determined which units need to be replaced, which need replacement of the motor and which need replacement of torque switches. JCO's were provided for units not totally qualified and replacement activity has been initiated.
Design Basis Accidents At the December 15, 1983 meeting your staff requested confirmation that all design-basis events at CR-3 which could result in a potentially harsh environment, including flooding outside containment were addressed in identifying safety-related electrical equipment at CR-3 which was to be environmentally qualified. Flooding outside the Reactor Building has been examined in GAI Report No. 1811, " Effects of High Energy Piping System Breaks Outside the Reactor Building". Flooding from external sources were evaluated in the FSAR. In the Auxiliary Building, modifications have been implemented to preclude the release of steam in the event of an auxiliary steam line rupture, thereby limiting harsh environment parameters in the Auxiliary Building to radiation.
Inside containment the environmental parameters were based on the design-basis postulated LOCA. Since the containment spray system at Crystal River Unit 3 . meets the single component failure l
criteria as required by Section 4.2.1 of the DOR guidelines, the LOCA parameters are considered more limiting than the design-
_ basis postulated HELB (per the DOR guidelines). To summarize, j., the flooding and environmental affects resulting from all j postulated design-basis accidents were considered in the I
identification of safety-related electrical equipment which was to be environmentally qualified.
4 4
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Ongoing Qualification Efforts Florida Power Corporation's discussions regarding the Qualification Program (Item III of Enclosure 2) illustrate FPC's sincerity in complying with the requirements of 10CFR 50.49 in a timely manner. We believe the environmental qualification doct? mentation maintained in the CR-3 Qualification File complies with the requirements of 10CFR 50.49. We also believe that CR-3 can continue to operate without undus risk to the public health and safety based on the JCos provided.
As discussed at the December 15, 1983 meeting, it is requested that supplementa SERs be issued to indicate that Florida Power Corporation's Qualification Program, as described herein, meets the requirements of 10CFR50.49 and that the deficiencies noted in the SER dated January 11, 1983 are considered resolved.
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Enclosure 1 _
The enclosed printout is a computer listing from the Qualification Maintenance data bank. TER equipment item numbers, categories, and deficiencies are identified as well as current resolution and planned modification numbers where applicable.
The listing is presented in the order provided in the TER.
JCO status has also been provided in the enclosed printout. The last column on the right indicates whether a JCO was proJided prior to the TER and accepted by the TER. The second to last column indicates whether JCO's were not required (NR) or required and submitted (YES) in the May 20 submittal. The staff indicated they did not require additional efforts on JCO's that had been previously accepted by the staff.
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<<<<<<<<< FLORIDA POWER CORPORATION >>>>>>>>>
CRYSTAL RIVER 3 OUALIFICATION NAiNTENANCE DATA T1ANM Mi FE-TER EQUIP. S(EW FME EQUIIMNT CtNTOENT fr# ERJIMNT 20 TER CMR NMR ID. N MER IESC./FiSE. CATEITRY KTICITEY KSfttfTION MRMMR Jt fl X0
-=====______________ ___________==._ .......~...._____________________________..~.............~...........m............~............m.....
07 2-43 CFV-12 MLW (MRAT(R !!. A I. DottKNTED EV!KNrE & RMLIFICAi!W Wttfr4M MS Cfm(IED 10 CtWIRM (10fENT 9W (RKR MRITRt $2 M-24 M VES VES MT!R 1RlWN 2. Emf!FtfMT VS TEST 91CITN AND ASSf(IATED Nmffir.TE MTA FCR T10Tms. DE WN00R MS LIMIT 0 roe 3. AC!NG KC#0ATICW EVAlllATION TEN CMTACTED TO CMJIPM TEST KTMT ArtilCAPILITY, FM
- 4. Mt!FIFD l!FE (R IEFtNEENT SOfDtf THIS E@f!FMNT, ifT4WMS IEE KWIRED AS TT RLr@fS TMT 9 9-000-2 7A. PEAK KNTMDM NTtKD R ImitRSE FOR TW EQUIFTENTS L((Afim. titTAtt
- 79. ITM FWSStM TO ImitM MTOR N@ limit #@ TORat SWITOf RFT1 Arf 4NT.
CONTAINENT 8. DEMICAL 9BAV IS[ tai!CW C7 2-44 UV-15 VALVE OFERATOR II.A 1. 00ClMNTED EV!KEE K OIALIFICAll(W WLKDC6M MS CO@l(IC 10 C(WIRM OMENT 9W (RKR Mt1ETRt 02 M 24 M VES VES MOTCR MIEN 2. ERJIINNT VS TEST SFICIMM NO ASS (TIATED NVff1 ATE MTA FGt M0yrRS. Tit Wi(0R MS LIMITCRGE 3. EING K0MOAi!(W EVAllMi!0N TEN CCWTACED TO CtWim EST KFTRI NT1I(tSilliY. FfR
- 4. OlWLIFKD LIFE OR RFtACETNT SOfittE THIS E0lflFMNT, (fTRAMS M KWlRID AS M KFMIS IMI 99 000-2 7A. MM TDfTRARRE NMKD R INAKRETE F@ Uf EWiffENTS LOCATl9f. (F9AK
- 79. ITAK FIES95E TO INfttM M0i(R NS LIMIT AND TORGE 91!T0f FIFINTENT.
CCWTAINENT 8. DEMIC4. NBAY ISCLAil0N 07 2-44 CTV-14 VALE OFTMinR ll. A 1. 00CtMNTED EVINNtY W RMLIFICATION MLED0lM MS CmRCTED TO Cfw!Rrf (tRENT 9W (RKR NMRS 82 05-24 4 VES 1TS MICR 1RlWN 2. EQUIFfENT VS TEST 91CIEN NO ASSnCIATED MEftAE IWTA FCR MOTmS. TE EN0f4 MS LIMIT (R0lE 3. AGING KCitAMTI(W EVMtmilm MN CONTACTED TO CtWIRM TEST REPORT NTtICAPILIIT. FOR
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- 79. ITE f1ESSIEE TO INfttM mifR AND LIM!i AND IfRGE SWIT0f Ki1NTffMT.
C(WTA!WENT 8. OfMICAL SFRAY ISftATI(W C3 2-74 FIN-14 VALVE (MMTOR II.A 2. EQUIFMNT VS IEST SITCIEN WLKirtM MS CONitCTED TO CCWIRff CtRTENT 9W (RIER WERS WM) MM mf(R LRIEN 3. MING llEGWMilCW EVAllMIIM N@ ASSOCIATED MEftATE ml4 FOR M0f(RS. IIE WNivW MS LIMITORRE 4. REIFIED LIFE (R EllACEENT SOf!4LE TEN CM1 ACTED TO C0KIFM FiSi ETYRT NYtIrAPillTV. All llENS NE C(WIKWD CtffED.
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<<<<<<<<< FI_ORIDA POWER CORPORATION >>>>>>>>>
CRYSTAL RIVER 3 CMNW. I F ICAT I ON MA I N T E HANCE DATA DANI' Mr FM-MR EGJIP. SEIV FR EWJIPENT OWfMNT MR E931 PENT 20 TER
WR M9WR IB. NMR KSE./Fitif. EATERRY KTIE!ENCY E9t ti![fti M@ l8MR JO .irg ww = u ___ _ _ _ _ _ _ , = = _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ __ _. _ _ _____
52 2-1944 9424-WT LEWL TIWNSMITTER N/A mi IWitKD IN Ewittlil(88 TO K IETALED IN01 IWVinlRY IEI4 LID). RS-14 AA 09 Im N/A IlosTIFIED IN TER AS SP-!A LT24.EEM AND ROTNtwi T-19 LIM.EE% KitML ASSIMD MMRS FOR itSI9t/INSTELAffres E AS 9nal, 115305 STEM RN. LEWL fndli(R 52 2-185A T-0&Ti LEiEL TRMSMITTEP II. A 1. 500KNTED EV!4WT T CIMIFICATI0li 70 K INSTALLFO (mi fwVtttRT IIIST4 LEP). (W la 4A-M Im Nm INNTIFKie IN KR AS T-IA4TM.EE5A NO ROSEN0ttfi T-lp LTM.%E% MTR ASSIMD WOMRS FtR KSiral/imiaLLAf!(es M AS 9 pat.
Il5E&A STEM EN. LEWL feeli(R
- 52 2-1954 T-026-8Pi LEWL TRM5MITTER II.A I. BonMNTED EVIKMT OF SRIFIEAil0N 10 K INSiaLLIO tmi FWVI0tf5tY INS 14 LED!. %10 A&M NR W
!NNTIFIED IN TER M T-lMT24.%EM Ne llDENDINT T-19 tT%34.E% KTUAL ASSIMD 18MERS FtR KSIGI/INStaLLA!!ON APE AS 9nse, ll531 EPA STEM EN. LEWL futIT(R 52 2-18 % SP-027-Fi LEWL iMN5MIIMP N/A MT INtitMB IN Ewitailfel TO K INSTALLEB (WT F1EVINISLY IElatLED). %IMM M N/A INNilFIED 14 TER AS SP-lHiM.EE54 AS R'TM40fi SP-194TM.%E% ACTM ASSI(MD N#MRS Ftm NS!(41/IETAllAi!91 M AS 99WI.
II53IMPA SEM MM. LEWL InliiCR 52 2-185A SP421FFT LEWL TRM9ffiRR N/A 180T III(ItMD IN EVAttailfst to K INStaLLE8 OtiliWVIMLY IETALIM. %In AAe
- FDiluff T-IP LT24. AE% FITM .%SIMD NMRS F(R MSI4f/IE14tLATi(et AE AS gnet, ll5309'A STEM MM. LDEL IRIIf(R F'Ae 4 x 41
<<<<<<<<< FLORIDA POWER CORPORATION >>>>>>>>>
CRYSTAL RIVER 3 OUAt_IFICATION HAINTENANCE DATA IWJK MY IM-WR EWIP. SCEN PAfE E9fifMNT C9f9thi NR EW!!MNT 20 ER NMR NMER ID. M450t MSC./FtWC. (AIEGWY MEICIDEY K9ttfil(UI NR WNfR J0 JO a _. ____ .._...__....__...____._...._..-......._..ws,.................,...
52 2-18 % T-029-Fi LEW1 iMN9' lier II. A I. IKl#ENTED EVINNrE & OIALIFICATION TO K IICIALED (Nnt FWytmlSLT IEfatED). Pl&MP NR W INNTIFKD IN RR E 9-lhti24.%EM ADO RD5EMUff SP-lO LI24.%E% KiiUIL ASS 100B N8906 F(R KSIm/INSTAttargree ME E grus.
l!5305PA STEM GEN. LIVEL HITOR 52 2-1854 T-033-tri LEWL TMN9 TITTER !!. A 1. DK10fMTED EVIEKT F QUALIFICAfine TO K INSTELED (*f INVISRI IETRLED). &lo M M Mt wt INM!i!ED IN TER AS T-letT28.%EM M ROE!Utif T-lO 1f24.2.44.% ACitK ASSIGED NMRS F(It MSIGE/IErumit(se M AS 9fue.
?!53D5PA SEM RN. LEWL pliOR 52 2-IBM SPirfi-Fi LEWL iMelTER N/A M)? IIELUIED IN EVRtElim 10 K INSTELED (mi NVIMISLT INSTAL 1ED). # t& M-dI5 Mt N/A INNitTIED 14 TER AS F-l&1724.34.EM A4 j
R05000fT F-lp ti24.%44.% ACilE ASS!fitD M990tS Im MSIGN/INSTALLATI(si AK AS Jful.
!!5305PA STEM MN. IIVEL N1817'It M 2-1954 T-072-IFT LIVEL TMN9tliTER N/A WT 18EttMD IN EvlittGTIft1 TO K INSTALLED (MOT fMV!GRY !EI4 LED). hio f4 M Mt N/A INNilT[ED IN I(R AS 7-14 LI2A.EEM AND RCElftf8T T-lO LT24 %4A.% ACitML ASSIGtD NtMRS FNt KSIGIIINSTRLATIfti AK AS 9ful.
Il5305PA STEM KN. LEWL plitR 53 2-115 lit-Col-til LEWL TMRitTER II. A 2. [Wf!!MNT VS TEST SFTCIKN TO K KitKEDI JfSTIFKAllfte FlR Cf9ff!WED NTMil0N MS P2M24 04 MS W
- 3. ACiss KOMMTION EVAttail01 KEN F90V!KD.
ROSOfteff 4. RutIFKD Lif E OR iEf1KtitNT SOfpt E ITER OfGrfT 9tli IKfWtETE-- JO HAD KTM IMVinRV
- 5. Ffl0GIAM FOR MIS ITG!AIETIrug r7Wittp.)
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CMIM AW IWilAllas I W .I's 4?
<<<<<<<<< FLORIDA POWER CORPORATION >>>>>>>>>
CRYSTAt_ RIVER 3 CMJAt_ I FI CAT I ON NA I N T ENAN('E DAT A BANt!
my flit-TIR Est!P. $[EN PArt EGulpf0ft C(vitteli TER Egl!! PENT 28 TER MNTR MNFR ID. NOWER KSC./Ft8C. CATEfVRY KTICKE) RSRl8TIMI PWR WMi# J0 #11 c __ _ _ _ _
W 2-115 K-#14T2 LEvil TRm911T!ER II. A 2. E0VIf1ENT VS TEST SFECIEN TO K EitKED: J611FRAll(40 FNt Cmi!WED WiMiliVI MS F2M 24-M VES m
- 3. MIE KGNNTIMI EWilmitel KEN F110Vi!TD.
flestinNT 4. R81LIFED LIFE (R EftACE!0ff EUEllitt (IER (EO(TF 9ffi IEftf1ETE- JO HAD KEN T9fvift4f
- 5. fil0mm F(R Ar.im KuleAT!rif fenVtyD. :
1852 IF 6. AGING SitetAlltel
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- 3. MIE KGMDATION EMlmiltsf fMTL !!52 Imi MS TO K flDtAITD. M TMeliitR m3 fl05ENUfi 4. SKIFIG LIFE (R lEFLNTENT SOERtf KitKED PRINT 2KFtEL 4 NITH TE IWKATED 194fL. ItWVER.
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54 2-175A PORV WI A(IELDUETER 81. A 1. RELIOffE2 EVIMMEE F OtMLIFICAi!91 INSTELED THIfa 70 C9FLEil0N W GletIFICAllftf EFT (Ri$ 10 INGE) fft Nil SAf ttTY 188EG 0737 (1MlliMNTS. IstS flEMLMD FIUt [O PWtCHR EIEEWO LIST KCALM N M SYSTEMS SMEfi(14RSIFKAffree tvi m3 St9TOCITLY KEN Pbi De Of FtR R.G.1.97 EFTGifS. Tit 2273m20 IIAN (utRS (Rut IS (tsptN11Y If4EING TO KHKVE A Ob1LIFIED ftmS & SATIFVING ITEM II.D.3 & NEG 0737.
55 2-:38 9P-IAt lO ELECT 9IC ISTOR ll. A 1. DOC 190fiED EVIKMT & Gut!FirATICII Sl95E010ff KVIEW ISICATED THIS E9J1fMNT MS irtATED IN A WWD 80t VES MILD ENVIRraffMT. IT E9f!! TENT IN THIS RR ITEM NOWER ELECTRIC MCHIltRY IMS MTN ApaEIVED AND IS ND LOMffR (91 TK EO M4TER LIST.
56 2-182 111F-34:30 ELECTRIC f0f0R II. A 1. DnftfENTED EVlftitY & QUILIFICATION 9.9T@ENT KVIEN INDICATED THIS E7JIFffMT MS LfEATED IN A (KKI E DES PILD EWif0ftNT. M ERl!MNT IN THIS RR ITEM MfMR ELECTRIC MCHMRY m9 KEN ARCHIVED AND IS 18) L981R ft1 T6E EO fMSTIR tISI.
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CRYSTAL' RIVER 3 OUALIIICATION MAINTENANCE DATA DANf(
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. Note 1: _
This -item (TER item 36) was discussed in detail at the December 15, 1983 meeting due to the significant amount of misunderstanding regarding the previous information submitted.
In the 1981 submittal,, FPC submitted SCEW sheets for SP-6A-PTl&2 and SP-6B-PT1&2 (SCEW page 2-186) and its anticipated replacement equipment, then identified as SP-6A-PT3&4 and SP-6B-PT3&4 (SCEW page 2-186A) . In that submittal SCEW page 2-186 was erroneously lined out although it accurately reflected installed equipment (Foxboro EllGH transmitters).
In reviewing the efforts of the TER, it is apparent that the anticipated replacement equipment identified on SCEW page 2-186A (Rosemount transmitters) was evaluated although this equipment had not yet been installed. This equipment is listed as Rosemount transmitters in TER equipment item.36.
When the equipment was walked down by a third party they were in fact identified as Foxboros (as listed on SCEW 2-186), but specific PT numbers were not identified. Therefore, the equipment was addressed in the May 20, 1983 submittal as SP-6A-PT3&4 and SP-6B-PT3&4 instead of by the actual tag numbers, SP-6A-PTl&2 and SP-6B-PTl&2.
All statements regarding the equipment however, remain the same.
The plant tag numbers submitted in the JCO in the May 20, 1983 submittal for this equipment should be SP-6A-PT1&2 and SP-6B-PT1&2.
After the discussion above was clear, the staff reviewed and accepted the JCO (enclosure 4) for the installed equipment.
Note 2:
This equipment is located in environmental zone 6. The only parameter considered harsh is radiation. The 40 year exposure at the equipment is 1 x 10 R. Total Integrated Dose (TID) does not 5
exceed 1 x 10 R until greater than five (5) days after the accident. TID for 40 years plus six (6) months post accident exposure is only 2.6 x 10 5 R and is. not expected to cause catastrophic failure of any insulating material. This equipment will be maintained in accordance with vendor recommendations.
Any additional aging evaluations are not warranted at this time.
This equipment is considered qualified for its current application and location. However, for operational considerations, this switch is being replaced with another switch.
Note 3:
The Crystal River Unit 3 Emergency Feedwater Pump was originally purchased as a non safety related item which for various reasons has become safety related. Because of its location, it has been evaluated to determine whether it can withstand the effects of a high energy line break. Although the motor is not located in a radiologically harsh environment, both the radiological and thermal susceptibilities of the motor's materials have been investigated.
l The motor is totally encased, water-air cooled and designed for a 70 C rise in temperature over a 40 C ambient temperature. A l walkdown has been performed which shows that the motor is not f subject to direct impingement and that the electrical connections are hermetically sealed. Analysis shows that the motor materials are appropriate to withstand the radiation and temperature j extremes which may be expected.
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The motor is tested, inspected and cleaned and dried annually.
Insulation resist,nce is checked and insulation is checked for cracks, peeling, non-adherence, discoloration or other signs of loss of physical integrity.
In the light of the above, FPC considers this motor qualified for its current application.
Note 4:
Conax Report IPS-769 " Feasibility Study Electric Penetration Qualification Crystal River - Unit 3" dated August 28, 1981, indicates the electric penetration assemblies themselves are acceptable but their terminations should be upgraded. Currently, the terminations are KULKA terminal blocks (see TER item 66) which are scheduled to be replaced with splices. JCO's have been provided for the kulka terminal blocks.
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Enclosure 2 .
AGENDA I. INTRODUCTION
-II. APPROACH FOR TER RESPONSE A. Address Identified Deficiencies by:
-Documentation
-Plant Walkdowns
-Maintenance / Surveillance Reviews
-Conservative Replacement Criteria B. Required JCO's III. QUALIFICATION PROGRAM A. Program Development B. Central File Restructure C. Data Bank Development D. Guide Specification Development E. SREF (Safety-Related Engineering Procedure)
Preparation F. Personnel Training IV. ISSUES AND RESOLUTIONS A. Generic B. Equipment Specific V.
SUMMARY
- A. Action Item Summary B. Clarification of Future Reporting Requirements j
L Enclosu'e r 3 METHODOLOGY TO IDENTIFY EQUlPMENT WITH THE SCOPE OF 10CFR 50.49 (b) (2) FCR CRYSTAL RIVER UNIT 3
- 1) A list.of all nonsafety-related (NSR) equipment located in harsh environments and connected to Class lE busses will be generated from Class lE electrical distribution drawings.
- 2) The e.ergency procedures will be reviewed to ensure that all equipment located in harsh environment and used by the operators appear on either the master list or the list generated by item 1 above. The item 1 list will be supplemented if additional equipment is identified.
- 3) The item 1 list will be reduced initially by determining which specific nonsafety circuits are tripped from Class lE busses by an accident signal or the operator during DBE's.
The documents used will be the flow diagrams, electrical drawings, FSAR chapters 7 and 8, and the emergency procedures.
- 4) The remaining equipment / circuits will be evaluated on a piece by piece basis. A failure modes and affect analysis will be utilized to determine if failure of the remaining NSR equipment could affect safety related (SR) equipment.
If not, no further action is required.
- 5) Circuit protection for NSR equipment whose failure would directly affect SR equipment will be evaluated next. If circuit protection is adequate no further action would be required.
Enclosure 3 (con't)
METHODOLOGY TO IDENTIFY EQUIPMENT WITH THE SCOPE OF 10CFR 50.49 (b) (2) FOR CRYSTAL RIVER UNIT 3
- 6) If circuit protec is inadequate, the effects of an electrical fault will 'etermined. If NSR equipment failure would influence SR equipment then this equipment must be qualified.
- 7) Any NSR equipment that must demonstrate qualification but doesn't, will be added to the safety listing, identified for replacement, and JCO's will be provided.
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I.
a Idencify all Non-Safety Related Electrical Equipment in harsh environment ~ and connected to Class lE bus l
N I
Identify all equipment'in -
ADD emergency procedures located in harsh environment but'
\
not on Master List or list' identified above Is equipment isolated from lE buses by accident signal or required operator No further , YES action?
action required NO
_ Would Equipment failure 40 further NO affect safety related action required equipment?
l YES No further Are there properly coordinated
' yg action required protective relays, circuit f
breakers, and fuses for electrical fault protection?
NO v
Equipment must a e degrade
,' YES i be qualified lE buses?
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vi NO l
f No further action
. required I. i i
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, Flow Chart for Identifying and Evaluating equipment within i the acope of 10 CPR 50.49 (b) (2) l l
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Enclosure 4 JUSTIFICATION FOR CONTIN ED OPERATION for-SP-6A-PT1, 2 and SP-6B-PT1, 2 Foxboro Transmitter
+
- 2) B&W- Proprietary Document 58-0079-001, " Test Data for Nuclear Transmitters" (Foxboro) FPC File F180-3TR-002.
I
- 3) Foxboro Test Report T2-1057, " Radiation Test of E-10 Series Differential Pressure Transmitter" (August, 1973).
- 4) Foxboro Test Report T3-1097, " Radiation Test of E-10 Series' Amplifiers -
Standard and Radiation Resistant l Types".
. Building Pressure vs. Time".
i
.SP-6A-PT3, 4 and 'SP-6B-PT3, 4 Foxboro transmitters will be replaced prior to November, 1985. The units will be replaced with Rosemount . transmitters, and will have documented evidence of qualification. The safety function of providing post accident monitoring information will be accomplished by l
pressure transmitters installed for the emrgency feedwater change. ,
In the interim between the present and scheduled replacement, the'following justifications for continued operation are given:
- 1. Figure 4-8 of the FPC 79-OlB Response (Reference 1) has been reproduced from Reference 2, " Test Data for Nuclear Transmitters". The test conditions shown in Figure 4-8 completely envelope the conditions specified (Reference 5) .
for Crystal River for high temperature and pressure condi-tions. Radiation conditions are discussed below.
- 2. Subsequent to the qualification tests performed for B&W (Reference 2) , Foxboro performed tests on E10 Series Transmitters and Amplifiers (Reference 3 and 4) to levels of 1.0 x 10 7 rads cr greater. Maximum error was 5.7% in zero shift for an amplifier irradiated to 2.2 x 10 8 rads (a factor of 37 above the 5.9 x 10 6 rads specified for Crystal River) . Maximum error at 1 x 10 7 rads was 4.2%.
Due to the fact that accuracy was better than specified for Crystal River (5%) for transmitters and amplifiers at 1.0 x 10 7 rads and the fact that maximum error at 2.2 x 10 8 was only 0.7% greater than specified, the transmitters should operate at the Crystal River specified radiation dose of 5.9 x 10 6 rads including margin.
(Note: These . transmitters are located in environmental Zone 39 and, as indicated by their SCEW sheet, are required for 24 hours post accident. Accumulated radiation doses 24 hours post accident have been calculated as 5.9 x 10 6 rads for environmental Zone 39).
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Enclosure 5 Referenced Portions of the May 20, 1983 Submittal (Referenced in Enclosure 1)
LUBRICANTS Title 10, Code of Federal Regulations, Part 50.49, provides the legal basis for the equipment to be considered as within the scope of the rule. In particular, paragraph (k) states that:
" Applicants for and holders of operating licenses are not required to requalify electrical equipment important to safety in accordance with the provisions of this section if the Commission has previously required qualification of that equipment in accordance with " Guidelines for Evaluating Environmental Qualification of Class lE Electrical Equipment in Operating Reactors," November 1979 (DOR Guidelines) . . . ."
Since Florida Power Corporation has been required to qualify the electrical equipment at CR3 in accordance with the referenced DOR Guidelines, and no explicit requirement for consideration of lubricants is specified therein, they are being addressed from two distinct approaches.
As part of the CR3 preventative maintenance program, Procedure PM-133 identifies the lubrication frequencies and the lubricants which are based on manufacturer's recommendations. Furthermore, lubricants, which are used in safety-related equipment, are purchased under the Safety-Related Catalog Purchase system using a Catalog Evaluation Sheet.
To supplement the above activity, an investigation of all lubricants used in Class lE equipment is ongoing; the aim of this study is to determine if the lubricants are susceptible to the respective environmental conditions. The results of the A-1
4 investigation will be a documented analysis of lubricants acceptable for- use in the specified environment and e- recommendations for replacement of those which are unacceptable for the specified conditions. The analysis results will be incorporated into the PM program..
In summary, it is the position of Florida Power Corporation that
~
lubricants should be . controlled as part of a preventative maintenance program which includes specifications based on electrical' equipment vendor recommendations and documented evidence of acceptability for the specified environment.
STATES TYPE NT TERMINAL BLOCKS Non-metallic materials used in the construction of Multi-Amp, States Co. terminal blocks are General Purpose Phenolic (Durez
- 791) and . Polypropylene (RTP 150); these materials have been evaluated for. their suitability in the specified environment for a desired service life ~ of 40 yaars, and the results of the evaluation'are summarized below.
The Micro-switch Engineering Report No. LTR-15027-1 submitted by l Hooker . Chemical, Durez Division contains an evaluation of Durez
- 791 material-temperature / life expectancy. Arrhenius methodology j was used to describe the temperature dependence of the velocity l coefficient of a chemical reaction to approximate the relationship between material life and temperature. It should be noted that parameters (time and temperature) were applied to establish failure criteria relative to flexual and impact strengths. Dielectric property of the material remained within acceptable levels without any failure. The approximate material
-life ' at ambient temperature is in excess of the 40 year plant life, and therefore the Durez #791 material exceeds performance requirements.
L A-2
. - , - _ . . _ , . . _ . . _ . . _ . . _ _ _ . . _ , _ . - , _ . _ . ~ _ - _ _ . . - _ . _ _ _ , . _ . _ - _ . - _ . -
The Micro-switch . Engineering Report LTR-15027-1 also provides radiation qualification test _ data for Phenolic materials. The materials _ tested were not exactly Durez #791, but they were of the same general purpose ~ compounds. Durez #791 is a general purpose phenolic filled with wood flour as the main ingredient.
The report demonstrated similarity of materials with respect to i comparable reactions to radiation. It was concluded that a material exposure to a TID of 1.3 x 10 8rads caused the Phenolic to be brittle, but did not affect the performance properties of the' component. A qualification program of Multi-Amp, States Co.
t included a test for terminal blocks to TID valuec of 2.2 x 10 8 rads, but no documentation is available at this time.
Heat stable Polypropylene material has a U.L. temperature index of 115 C (Allen-Bradley Co. , Bulletin 798 Control Center -Serial No. 967929 Non-Metallic Component List). The U.L. temperature
! index is considered the maximum temperature at which the material
'c a n be used continuously. An article ("A New Temperature Index:
Who Needs It"), published September 1970 in Modern Plastics j discusses the index and-how it was established. The temperature index is the point where the property of impact strength, tensile strength - and dielectric strength is reduced to one half of its l new value at the conclusion of 5 x 10 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Using the 10 C
_ rule, it was concluded that the approximate life of the Polypropylene material is in excess of 40 years.
Based on information provided in the document by J. F. Kirsher and R. E. Bowman, " Effects of Radiation on Materials and l Components, Reinholf Publishing Corp. (1964), radiation threshold values for polyethylene, which are believed to be comparable or lower than those of Polypropylene, are in excess of 1.7 x 10 rads for the properties which may be of importance for the particular application in terminal blocks.
A-3
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In summary, the non-metallic materials used in the construction of Multi-Amp, States Co. terminal blocMti were evaluated for their suitability in the specified environment for a desired service life of 40 years. The States Co. conducted qualification tests to meet requirements of IEEE 323-1974, but no data or test results are known or published. Several independent analyses were conducted to determine the equipment qualification and it is expected that these results, when released, would only confirm the results of these analyses. Information provided by the manufacturer indicates that the terminal blocks were successfully tested under postulated environmental conditions of DBE.
Terminal blocks in several locations at CR3 (SW-6 & 14; WD-3) will be exposed to higher radiation doses during the equipment service life than the radiation threshold value of Polypropylene material. These terminal blocks are scheduled for relocation to a less hostile environment prior to November, 1985.
l A-4
.7USTIFICATION FOR CONTINUED OPERATION for BS-1-dPT 1 & 2 Bailey Meter Co. Flow Transmitters TER Item 45 1
- References 1). FPC IE Bulletin 79-OlB Response, Figure 4-7.
2). B&W Proprietary -Document 58-0081-00, " Type Test Report of Bailey Meter By Differential Pressure Transmitter", dated 3/12/73.
The. building spray flow transmitters will be replaced prior to November , -1985. The units will be replaced with transmitters having documented evidence of qualification, or other methods will.be specified to reduce the harsh environment to acceptable levels. SCEW theets for the new transmitters will be completed upon the receipt of the new transmitters and associated test reports.
In ' the interim between the present time and corrective action,
. the following justification is given:
- 1. The only harsh environment for this transmitter is radiation 5
at a level of 2.6 x 10 rads. Reference 2 indicates quali-4 fication to greater than 2 x 10 rads.
i i
l 2. Since the building spray transmitters are required only to i detect the onset of spray during the first few minutes of
( the accident, total integrated dose will be below the quali-l fied radiation dose at the onset of building spray.
I l 3. Failure of the transmitters can be tolerated later in the accident sequence since spray can be detected using other 1
i instruments, such as pump operating indicators, pump outlet / inlet pressure, etc.
A-47 l
- l. , , . _ _ . _ _ _ _ . . _ . - _ . . _ , . _ _ . . . _ ___- . _ . . _ . . - . . . _ , . , . , . . _ _ , _ . . . _ , _ , _ , _ . _ _ .
Justification for Continued Operation DHV-5, DHV-6; DHV-34, DHV-35 Limitorque Motorized Valve Operators
~
TER Item 9;2 References
- 1) FD-302-64, Revision 27, Decay Heat System Flow Diagram
- 2) FD-302-651, Revision 25, Reactor Coolant Flow Diagram
- 3) FD-302-702, Revision 12, Core Flooding System Flow Diagram
- 4) Environmental and Seismic Qualification Guide Specifications and Data, SP-5095, dated April 29, 1983; Section 4,
' Environmental Qualification Data'
Background
. During the TER response activities, four (4) Decay Heat Valves (DHV- 5, 6, 34 and 35) posed a potential problem. Walkdown results indicated the motors on these actuators contained class H insulation which has not been tested to high radiation doses.
- _ Hence, the radiation resistance of those motors is undocu-l mented. (The calculated total integrated dose (TID) for 40 years plus a design base accident for the installed location was obtained from reference 4 and included herein as Attachment 1.)
l To eliminate any question of radiation resistance, Florida Power l Corporation prudently decided to replace the motors with class B or RH insulated motors and remove the class H insulated motors l due to inconclusive documentation.
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L l A-76
'A
' Actions _
i v.
Florida Power Corporation initiated #efforts to replace the motors s
on these actuators, duEing ther Spring 1983 outage, as well as pursue . justifications f5r- continued- operation in the event replacement ' equipment could.- not 'be procured prior to the end of the outage.-
JCO Efforts Rev'iew of References 1 through 3 revealed there were no redundant system. arguments to provide a sound JCO. Material breakdowns for class H. insulation have not been-made available hence, similarity l in composition arguments are not feasible. Since radiation exposure was the only item of concern, a justification for continued ope $ation on 'the bas,is, of operating experience of similar equipment in Voperating plants was pursued. ,
Such a JCO would be consi'dered vali'd if:
- j i . '
- 1. The actuators in' the operating plants contained motors with class H insulation,'and
, 2. The actuators' and motors ,in the operating plants had 4 g received a TID wnich.excee'ds: '
')
'a. the current TID of the motors installed at CR-3 r 'olus l
l
/ b. .the KID the installed motors will receive during a '
s design base accident olus 3 '; ,
l c >
l
- c. the TID that will be received between the present l ,
l and the pla.nned date of replacement.
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! .First, calculate the dose we must exceed.
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- - /I ) . +.k, >
The equipment is located'in zone 36. Attachment l indicates a 40 8 6 year dose of 1X 10 R or a yearly dose of 2.5 X 10 R for this zone. The scow sheets for this equipment . indicate a required
. post accident operating time of 30 days. Attachment 1 indicates 4
a dose of 8.9 X 10 R for the associated post' accident operating 1 time.
Crystal River maintenance records indicate these motors are the 7
originals,.therefore the current TID is 1.25 X 10 R.
The dose received for the accident is 8.9 X 10 R.
. Assuming the motors remain in place until the next refueling
- . outage the additional dose received will be 1. 5 X 2.5 X 10 R =
3.75 X 10 R.
. Therefore the total dose received by a class H insulated motor in
-other operating plants must exceed:
7 1.25 X 10 R
+ .009 X 10 R 7
+ .375 X 10 R Total 1.634 X 10 R Utilizing the EPRI Equipment Qualification Data Bank, plants that use the same type equipment were identified. Plants in operation L longer -than CR-3 whose radiation specification was equal to or greater-than CR-3 for the equipment of concern were contacted.
It was learned that an operating BWR had similar equipment i 8
- located in an environment which would receive a TID of 2X 10 R l in 40 years. The current TID of the similar equipment is 7 X 5 X l 6 7 10 R = 3.5 X 10 R. This exceeds the anticipated dose calculated for CR-3 and supports its continued operation.
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A-76-6 l
Realizing that the larger dose rates occur during operation a comparison of the operating reports for both plants was made to validate implicit assumptions (e . g . , yearly operation so that the yearly dose is indeed rcceived).
The BWR's number of hours critical exceeded those of CR-3 by 10,725 (as of 12/1/82). Cumulative unit availability factors were comparable.
Based on the above results the continued operation of CR-3 is justified on the basis of operating experience with similar equipment in harsher environments at other operating plants.
A-76-c
DISCUSSION ON DWV-160 Record 0052 Vendor correspondence (L200-3VC-007) indicates Test Report B0003 (L200-3TR-003) is applicable to this valve motor operator.
This valve is located outside containment. Its only safety function is containment isolation in the event of a LOCA or HELB inside contal maent. While the valve must function during the accident, it is not exposed to accident environment. The environment at the equipment's location does not change during or following the accident for which it must function.
Therefore this equipment is not considered to be within the scope of 10 CFR 50.49. This equipment will be removed'from the Master List.
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DISCUSSIOt3 ON EFV-3/4; EFV-7 & 8 Records 0055; 0056 Vendor correspondence (L200-3VC-005) indicates that Test Report B0003 (L200-3TR-003) applies to the equipment listed. Thus, the only remaining concern is that testing does not completely
< envelop the required accident profile. The accident profile has an initial temperature spike that returns below the testing profile within 16 seconds. Because of the extremely short duration of this temperature peak, the equipment internals will not realize the higher initial temperatures. Thus the temperature peak is not an actual concern and the equipment is considered qualified for its current application.
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DISCUSSION ON Continental Wire & Cable Co.
Instrument and Thermocouple Extension Cables Record 0099 A review of FPC Purchase Order No. PR3-2178 was conducted, along with referenced bills of materials and their cable descriptions, to identify type and insulation materials of instrument and thermocouple cables; as a result of this review it can be summarized that all cables have silicone rubber insulation type CC-2193. The outer jackets are either glass braid or silicone rubber type CC-2193. F-C2935 applies to cables having glass- .
braid outer coverings. Anaconda Co. No 79118 applies to cables with silicone rubber outer jackets. Available qualification test reports Anaconda Co. No 79118 and Franklin Institute No. F-C2935, and its addendum dated November 1970, provide qualitication test data for previously mentioned materials on comparable test samples to specified cables. Individual test reports alone provide only partial qualification data. For this reason, the combination of both reports and their test results was used for evaluation of cable performance in the CR3 application; this approach can be justified, since the insulation material is the same for both tested samples and specified cables.
Since the CR3 cables have a specified operability time of six months, it is necessary to compute a post-accident time p9riod, for which the cables are qualified, for comparison to the post-accident operability time required.
The approach used was to calculate from the post-LOCA l j!
time / temperature profile, an equivalent time / temperature q . .d W.
combination with the temperature being of the postulated, CR3 U7 f; f, post-accident environment. The transformation of a given time '%ph :
and temperature to a longer time period at ambient temperature M; ~
- 3. . n was based on the slope of the Arrhenius plot provided by the 4 ] / r1 .
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manufacturer; this approach necessitated certain assumptions, namely:
- 1. The Arrhenius line for the pre-accident thermal aging can be applied at a post-accident time / temperature combinatin It is assumed that the slope of the line can be used ann .ere on the log t vs. 1/T plot. The calculation of a slope <as included.
- 2. The post-LOCA temperature remains constant at 150 F after 5 seconds for the duration of the operating time of six 10 months; this is a conservative assumption, since some decay of the temperature below 150 F is expected.
- 3. The post-accidene environmental pressure decays close to atmospheric and does not affect the calculation.
The CR3 temperature peak during LOCA is above 280 0F for lass than one hour. The silicone rubber cable was tested to a temperature of 280 F for a total of three hours in the Anaconda test. A cable insulated with the identical silicone rubber insulation (CC-2193) was tested in F-C2935 to a level of 3400 F for two hours. The only component which can be presumed to fail, therefore, is the outer jacket which was different in the two referenced tests. The jacket provides physical protection only and, therefore, can fail without compromising the safety-related function of the cable. Pressure differences will not affect the ability of cables to carry signals or current. In addition, F-C2935 indicates a pressure transient which is greater than CR3. The cable is considered qualified since its safety-related function has been demonstrated in environments which exceed LOCA environments at CR3.
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l l
l DISCUSSION ON Kerite Co. Termination Procedure #39-69 Record No. 0106 A letter from the Kerite Co. in the FPC files states that the FIRL Report F-C4020-2 applies to termination procedure 539-69.
According to the vendor:
"Kerite Tape Type Terminals are qualified by comparison with qualified splices using similar designs and identical materials. The 39-69 terminal corresponds to Splice 38-69 covered in FIRL Report F-C4020-2...
"The following is a listing of Kerite Kit materials and corresponding manufacturer designations:
Kerite Kit Manufacturer Designation i Bishop Bi-Seal 3 Bishop Bi-Seal 3 (see Section 1 for comments on this tape) .
Bishop W962 Bishop W962 Kerite Friction Tape Kerite Friction Tape Glass Electrical Tape Permacel P212 Glass Electrical Tape (not Scotch 27)
Silicone Rubber Tape Scotch 70 Silicone Tape "The Bishop Bi-Seal 3 tape has been subjected to radiaton and maintains its physical properties up to 0.1 megarads of exposure.
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"The purpose of both terminals is to provide ~a lug aeal to prevent the cable from breathing during normal thermal cycling allowing moisture to collect in the conductor strands. The Bishop W962 tape will maintain its integrity during 40 years of normal service (both thermal and radiation aging) and a LOCA accident. The Bishcp Bi-Seal 3, as stated in earlier conversations, starts to revert at radiation levels above 1 megarad (0.1 includes safety factor).
Therefore, its performance cannot be guaranteed at levels above this, but our judgement is that it will still adequately seal the end of the cable."
Based on the above information, it is the position of FPC that the subject equipment is qualified for its current application.
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DISCUSSION ON FWV-29, 30; FWV-31, 32 RECORD 0063; 0064 Vendor correspondence (L200-3VC-007) indicates that Test Report B0003 (L200-3TR-003) applies to the equipment listed. Thus, the only remaining concern is that testing does not completely envelop the required accident profile. The accident profile has an initial temperature spike that returns below the testing profile within 16 seconds. Because of the extremely short duration of this temperature peak, the equipment internals will not realize the higher intitial temperatures. Thus the tamperature peak is not an actual concern and the equipment is considered qualified for its current application.
'4 A-128
DISCUSSION ON '
RC-4A-TEl & 4; RC-4B-TEl & 4; RC-5A-TEl, 2, 3& 4; RC-5B-TEl, 2, 3& 4 Rosemount RTDs References r
- 1) B&W Report Number 58-0372-01, " Qualification Test Report -
Rosemount 177HW Sensor, Temperature, Resistance Type",
dated January 19, 1978. ,
=
Florida Power Corporation disagrees with the Franklin Research Center (FRC) evaluation of the Rosemount RTD reactor coolant -
temperature sensors.
The sensor, thermowell, mounting nuts and gaskets are subjected to temperature and pressure conditions during normal operation which exceed harsh environment parameters. Reactor coolant _.
temperature exceeds 500 F and pressure exceeds 2000 psi during normal operation. Testing of these components at levels of approximately 300 F and 65 psig is unnecessary, since normal operating conditions far exceed accident conditions. It should be noted that pressure testing to 3750 psig was performed on the thermowell, and calibration at 610 F was performed on the sensor as pre-qualification for the RTD shewing capability at environments which are more severe than accident environments.
w_ _-
Two RTDs with connection heads attached were subjected to 3.8 x 10 8 rads and 3.0 x 10 8 rads without failure. Connection head capability during LOCA conditions was demonstrated by similarity to a unit tested to a maximum 350 F, 67 psia environment. The tested unit had a painted outer surface versus a bare surface for the analyzed Model 177 series. Although the qualification tests A-225 y&gg-->..
N W.
EE
were performed on the 177 HW model, the differences between the 177 HW and 177 GY were analyzed to show applicability of the test reports to the 177 HW model used at Crystal River.
Based on the above evidence of qualification, FPC has concluded that the Rosemount RTDs are qualified for their current applica-tion.
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