ML20003C802

Forwards Response to Addl Data Requests of Suppl 3 to IE Bulletin 79-01B, Environ Qualification of Class IE Equipment. Util Does Not Plan to Test or Replace nonsafety- Grade Equipment

ML20003C802
Person / Time
Site:	Brunswick, Robinson
Issue date:	01/30/1981
From:	Furr B CAROLINA POWER & LIGHT CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20003C800	List: ML20003C799 ML20003C802
References
IEB-79-01B, IEB-79-1B, NO-80-121, NUDOCS 8103180425
Download: ML20003C802 (18)

Text

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7-January 30, 1981 FILE: NG-3513 (R; SERIAL: NO-81-121 Mr. James P. O'Reilly, Director United States Nuclear Regulatory Commission Region II 101 ':arietta Street, Suite 3100 Atlanta, GA 30303 H. P. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 LICENSE NO. DPR-23 REVISION 3 TO IE BULLETIN 79-01B NINETY-DAY REPORT ~

Dear Mr. O'Reilly:

Attached you will find our response to the additional data requests of Supplement 3 to the subject IE Bulletin. Qualification data is provided for installed equipment resulting from TMI Action Plan requirements and a listing of equipment used to achieve a cold shutdown condition is also supplied. Cold shutdown equipment which is safety related and therefore has already been reported upon in previous sutaittals is indicated fa the attached list.

We were pleased to note the additional guidance provided by Mr. Eisenhut's letter of Jan uary 19, 1981, "Information Regarding the Program for Environmental Qualification of Safety-Related Electrical Equipment (Generic Letter 51-05)".

In accoruance with the guidance provided in that letter, CP&L has sapplied that qualification data for non-safety grade cold shutdowr equipmenc which is readily available. Since this equipment is non-safety grade and has no qualification requirements, CP&L has no plans at this time to test or replace any of the non-safety grade equipment.

810318 0%6

Mr. James P. O'Reilly January 30, 1981 If you hdve any questions, please do not. hesitate to coatact my staff.

Yours very truly, g% V. .

. $0TARy i ... i .

B. 3. Furr c ~

Vice President I F/jgg( Nuclear Operations n

.C..... . . .h..

Attachment Sworn to and subscribed before me thist fC day of G m ,,,m, 1980.

nnD Y M

)fotary Public My commission expires /d/ // 9 !/f

/

cc: Mr. Don Neighbors (ONRR)

Mr. Zoltan Rosztoczy (0NRR) l l

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8 SYSTEM: AUTOMATIC DEPRESSURIZXrION COMPONESTS

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pm7 PRIMA RY IDENTIFICATION ^ '

NUMBER GENERIC NAME NS EbhTSIDE B21-FT-4157 Sensor X B21-FT-4158 Sensor X B21-FT-4159 Sensor X B21-FT-4160 Sensor X B21-FT 4161 Sensor lX B21-FT-4162 Sensor X B21-FT-4163 Sensor X l

B21-FT-4164 3ensor X B 21-FT-4'.65 Sensor I X B21-FT-4166 Sensor , X B21-FT-4167 Ses2or X B21-FAT-4157 Preamplifier X B21-FAT-4158 Preamplifier X B21-FAT-4159 Preamplifier X B21-FAT-4160 . Preamplifier X

B21-FAT-4161 Preamplifier , X B21-FAT-4162 Preamplifier X B21-FAT-4163 Preamplifier X B21-FAT-4164. Preamplifier X B21-FAT-4165 Preamplifier fX 321-FAT-4166 Preamplifier l X B21-FAT-4167 Preamplifier , , X Amphenol Coaxial Connector X f 1 i

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SYSTEM: COMMON COMPONENTS

• COMPO:.'E!.TS

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PIA!.T PRIMARY IDE .TIFIC' T10N CO:.7AI:.r. 7 h"JM3ER E00"IION GENERIC NAME

'INSIDE bCTSIDE X-100E Electrical Penetration. X X X-100H Electrical Penetration X  :(

YL20 Instrument Cable X I

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ENVIRONMENT DOCUMENTATION QUALIFI- Otfr-CATION STANDING

• I.QUllt!ENT DESCRIPTION QtlALI FI- METil0D ITt:MS PARAMETl.F SPECIFI- QUALIFI Sl'E C I F I - CATION CATION CATION CATION SYSTEM:-Automatic Depressurization OPEllATI!K

• TIME Short PLANT ID. NO. B21-FT-4157, 4158, 4159, 4160, 4161, 4162, 4163, 4164, 4165, 41t,7 COMPOtENT: TEMPERA- Profile

• Sensor FSAR TUP.E A

( F) (Attach.1; ttANLFACT URE : NDT International Profile ,

PRESSURE MODEL NO. ries 700 (PSIA) (Attach.1 4

RELATIVE FSAR ,

llUMIDITY 100 FUNCTION: Flow Detection -

(%)

CllEMICAL *

ACCURACY

N/A SPEC: N - -

SPRAY -- --

DEMON:

l (See SERVICE: Automatic Depressurization Valve FSAR RADIATION Attach.1) *

(RADS) 1.1 :: 108 LOCA110N: Inside Containment

• i AGING FLOOD LEVEL. ELEV.: N/A SUBMERG.

ABOVE FLOOD !.EVEL: YES N/A N/A None NO SAF TY CATEGORY: Essential Active

• See Equipment Status Sheet Attached i

ENVIPONMENT DOCUMENTATION QUALIF1- OlTr-CATION STANDING .

I.f(I P:"r NT DESCR IP f !ON ME'I tt0D ITI:MS PAi?AMETES SPECIFI- QUALIFt.4PECIF1 CATION I QUAI.lF1-CATION CATION CATIO1 SYS1tM: Automatic Depressurization opg.itAT I Nt

• TIME Short PLANI ID  :.o. B21-FAT-4157, 4158, 4159, 4160, 4161, 4162, 4163, 4164, 4165, 4166, 4167 Profile

• COMPONENT: TEMPERA-Preamplifier E .

TURE Pf) (A t tach .1 )

MAN UF ACTURI. . NDT International, Inc.

PRESSURE 400A (PSIA)

MODEl. NO.

1 tlELAT IVE '

• HUMIDITY 100 FUNCTION: Signal Amplification

( *4)

GilEMICAL h ACCURACY: N/A SPEC:

SPRAY DEMON:

(38"

• SERVICE- Safety I elief Valve Monitoring RADIATION Attach. 1 )

System (RADS) 1.1 x 10 8 1.0 CAT I ON : Outsioe Containment RX 20 SCEN AGING FLOOD 1.EVEL ELEV . : N/A SUBMERC.

N/A N/A None AHOVE FLOOD LEVEL.- YES NO SAFETY CATILORY: Essential Active

• See Equipment Status Sheet Attached

1 ENVIRONME!1T DOCUMErlTATION QIIALIFI- OUT-HE FEREf1CE CATION STANDING Eqtfil'MFNT DESCRIPTION QllAI.lfi- METilOD ITLMS PAllAMETEP SIECIFI- QUALIFI- ' PE C I F 1 -

CATION CATION CATION

_ CATION SYSTEll: Common Component OPEllAT i fH 4 Analysis None TIME :ihor t 30 Days 54,13 Simul.

PLANT ID. NO. X-100E .g. es t X-100H A A A Simul.

COMPONENT: Electrical Penetration TEMPERA-None Assembly TURE Attach.1 W FSAR 13 Test MANUFACTLfRE Westinghouse PitESSURE B A A Simul.

XX FSAR 13 Test Hono Clas's E Penetration (PSIA) Attach.1 MolsEl. NO.

HELATIVE A Simul.

FSAR None H UN I DITY 100 10'0 13 Test FUNCTION: Containment. Building

( 7.)

CilEMICAL ACCURACY: N/A SPEC: None SPRAY DEMON:

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A SEltV ICE : SRV Monitoring FSAR 'fone ttADI AT lufi 1.1 x 10 0 L.256x10 8 2 (RADS) _.

l.OCA l lOf t: Inside/Outside Containment AGING FLOOD LEVdl ELEV.- N/A SUBMERG.

N/A N/A None ABOVE Fl.00D 1.EVEL YES NO SAi ETY CATEGORY: Escential Passive A :ubmitted with 90-Day Report, Rev. 1, dated 10/31/80

N DOC IENTATION QU AI.I F I- OUT-ENV I RONME N'1 N STANDING CATION lI ITEMS

! Qifli?!ENT DESCRIPTION PARRfETE14 SPECIFI- yUA! IFI. iPECI CAT NN FI- QU Al.I F 1- PETil0D CATION CATION.., CATION SYSTEM: Common Components OPERATt!R

• TIME Short PLANT II). NR General A

• COMP 0i4ENT: Connector TEMPERA-FSAR TURE op) Attach.1 MAN U F ACTIIRI'. Amphenol B

PRESSURE

• FSAR (PSIA) Attach.1 MODEl. NO. 36500 (Jack) 34500 (Plug REl.AT IVE

• 100 FSAR 11UMI DITY FUNCTION: Calle Connection

( *4)

CHEMICAL ACCURACY: N/A SPEC: ,

SPRAY DEMON:

SERVICE: SRV Monitoricg RADIAFION 1.1x10 8

• FS'AR

(,g gs )

Inside/ _

l.0 CAT 10N:

Outside None AGING FLOOD I.EVEf. I: LEV. N/A SUBMERG.

N/A N/A  %

ABOVE FLOOD I.EVEL YES NO SAf t:fY CATE).ORY: Essential Passive

• See Equipment Status sheet Attached

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DOctiMEtiT AT IOtl QUAI. l F 1- OUT-1.NV i RO illi til -

NCE. CATinti STANDING ITEMS I:4111PMINf it.SChIPTlot! QU AI. ! F I- ME'ifloD P ARAttETI d ':s>EC I F 1- QUAL.1F1 iPECIF1- CATION C AT 10!1 CATIOtt CAIION ,

OPER AT I N(

A SYSTEM: Conunon Components A AI T T NME TIE D YS 51 YL 20 g,,n ,,

Pl. ANT ID. NO. -_.

12 Pair #20 TEMPERA,?rofile & A A sit 10L, COtiPollENT: E .

41 22 TEST t10NE TilRl gop) [ Attach.1 MAEllFACTilRE: lioston Insulated Wire & A SIMUI.,

Cable Cc l'RESSliRE 41 22 TEST NONE 14,9 Z (PSIA)

MODEl. NO.

REl,AT IVE g Ms gIgg L, I 22 TEST t10NE N"" 100 100 41 FUNCT IOti: Instrument Cable (7 )

CIIEM' CAI' BORIC ACCURACY : il/A SPEC: 22 S I Mu l..

SEKAY ACID - 140NE DEMON: tao TEST A

instr;nentation A SEQU Etl .

NONE SERVICE 40 22 TEST

,  ;) lx10 lxlO

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1.OCAT ION: Out t.ide Containment

• VII-A AGING '

N/A SullMERG. NONE FLOOD 1.EVEl. ELE 9 N/A YES N/A ABOVE Fl.OOD I.EVEI.'

t10 Safety Category: Essential Passive Rev. 1, dated 10/31/80 g Submitted with 90-Day Report,

.._ = _ _ - . _ .. -

EOUI?ENT STATUS The components of the Safety Relief Valve (SRV) Monitoring System as installed per NUREG 0578 is currently undergoing an environ = ental qualification test (pe- Terr 323-1974 standards). The test is being performed at Wyle laboratories, Huntsville, Alabarus (Wyle's Job No.

4 54098) and is expected to be completed by August 1901. If any ecm-ponents are found not to be qualified, the NRC will be infor=ed by the appropriate reporting mechanism.

We have not been able to find qualification data for the A= phenol 4

connecter They will be replaced by connectors manufactured by Gul::en Industries wh2.ch are included in !te ongoing test at Wyle 1,aboratories..

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ATTACHMENT 1 DETERMINATION OF ENVIRONMESTAL SERVICE f MOITIONS

1. GENERAL Chemical sprays are not part of the BSEP design basis. Demineralized water sprays are used inside Containment and have been considered. [

In the design of the BSEP pressure suppression contaic=ent (i.e. drywell-torus) flooding is not considered to be a credible accident environ =ent, due to the high volume, low hydraulic resistance flow paths frca the dry-well to the torus which imediately direct LOCA blowdown flow away from the drywell.

2. INSIDE__ PRIMARY C0hiAINMENT l- HEL3 does not affect environmental conditions inside the Contain=ent and .

therefore is not considered for in-containment primary components.

i The drywell and torus environmental conditions following a postulated LOCA are discussed in FSAR Section M7.9 and Design Report 12. Te=perature and pressure response curves are labeled Profiles A and B respectively and are attached.

FSAR Figure M7.9' also provides values for post accident integrated radia- "

tion dose to-equipment in the Primary -Containment and these have been

-used in our evaluation.

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3. INSIDE SECONDARY CONTAINMENT a.- Radiation Post-LOCA radiation levels in various regions of the Reactor "uilding

, are due to leakages from the drywell and due to fluids which are re-circulated from inside the Primary Containment to accomplish long-term cooling following a LOCA. The LOCA-induced radiation environment, and the subsequent transport process, is evaluated in accardance with

. the guidelines in NUREG-0588 as follows:

fi) 100-percent of the noble gases, 25-percent of the halogens and 1-percent of the solid fission products are assumed to be air-borne in the drywell.

2) 50-percent of the halogens and 1-percent of the solid fission i- products are assumed to be in the suppression pool. water.

3) .Drywell leakage is assumed to be 0.635-percent per day for the duration of' the accident.

4)' ' For conservatism, all leakages from drywell are assumed to be leaking into any one compartment adjacent to drywell. The leakage outflow from that compartment is then assumed. to be uniformly Emixed in the remainder of the Reactor Building. The conservatism

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.of this ' method lies in the conservatively high compuced values for.the concentrations of:fissi:n products in each of the ESF

. compartments.

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l The resultant peak activities in various regions of the Reactor Building are presented in Table IV-1; the time-integrated doses fcr the ESF components in Reactor Building are presented in Table IV-2,

. and as referenced in BSEP Design Report No. 12 of March 1972.

. b. Temperature' Our continuing evaluation of the Reactor Building post-LOCA environ-

. ment disclosed that for elevations 20-feet and above the curve used in the 45-Day Report did not consider all heat sources.

I The temperature profiles for the Reactor Building post-LOCA and post-HELB are attached for El. 20' to 117' as Profile E.

Detailed qualification data has not been established as yet for the SRV position indicating equipment, however we expect type testing now underway will provide this data.

. c. Pressure ReactorL Building peak pressure data is included as an Accident Pressure Peak of 14.9 PSIA for El. 20' and above. This peak is applicable only

-during the HELB and is of such short durction that the effect upon Y

-essential equipment is con 4Ldered to be negligible, and does not affect'itn. operability. For post-LOCA conditions there is no sig-  ;

nificant pressure increase in the Reactor Building.

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d. Ilumidity The bounding post-accident humidity in the Reactor Building is con-sidered to be 1007. (HELB) .

Significant flooding doas not occur outside the drywell as a result of the postulated HELBs.

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TABLE IV-1 LOCATION OF SUBSTARIAL FISSION PRODUCT INVENTORY APPROXIMATE PEAK SOURCE LOCATION INVEEORY. Ci

' Reactor Building (Uniform Dispersal) Total: 8.1 x 10 6 S.G.: 5.6 x 10 6 Iodine: 2.5 x 106 SGTS Filters in Reactor Building Iodine: 4. 2 x 105 Drywell Total: 5.5 x 108 N.G.: 4.0 x 10 8 Iodine: 1.5 x 10 8 Pressure Suppression Chamber Iodine: 3.0 x 10 8 Typical Small' Compartment' (Adjacent to drywell in Reactor Building) Total: 10 5

- Typical Large Compartment (Adjacent to drywell in React.or huilding) Total: 10 6 l

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TABIE IV-2 .

EXPECTED DOSES FOR RADIATION SENSITIVE ESF COMPONENTS Normal .

Operating Maximum Accident Doso Dose Threshold (Rads) (Rads)

Component Location (Rads) 40 Years First llour 30 Days Comments Limitorque Inside 5 x 10 7 6 x.10 7 --- 4 x 10 7 Threshold dose applies to lubri-Volve Operators Drywell (Maximum) cation seals, made of Viton.

These seals can be periodically replaced to keep the normal oper-ating dose substantially below the threshold. The listed normal operating dose is the maximum dry-well dose, i.e., at the outside surface of the sacrificial shield at core mid-plane. Laakage from the seals does not imply failure of the operators to function.

. Solenoid Reactor 6 x 107 2 x 103 1 x 10 5 9 x 10 6 Accident dose based on conserva-

- Vs1ves Building tive assumption of non-uniform Below 20' E1. mixing of fission products in Reactor Building. Dose given is maximum for any compartment adja-cent to the drywell. Normal operating dose based on 5 mr/hr for 40-years.

Pressure Reactor 2 x 10 8 2 x 103 1 x 19 5 9 x 10 6 Same comment as for solenoid Switches / Building valves.

Transmitters below'20' E1.

Pump Motors: RilR Room o' Lube -CS Room 1 x 10 8 2 x 103 1 x 105 9 x to 6 7

o Insulation 5 x 10 2 x 103 1 x 105 9 x 106 o Seals 5-x 10 6 2 x 103 1 x 105 9 x 106 Failure of seals leads to leakage, ,

which does not render pump in-operable.

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TABLE IV-2

-(Cont'd.) Pago'2 of 3 .

Normal -

• Operating bkximum Accident Dose .

Dose Threshold (Rads) (Rads)

Component Location (Rads) 40 Years First flour 30 Days Comments

' Electronics: Reactor

.o Semiconductors Building. 1 x 108 2 x 10 3 1 x 195 9 x to 6 o Capacitors Below 20' El. 3 x 108 2 x 103 1 x 105 9 x to6 o Inductors 2 x 109 2 x 103- 1 x 105 9 x 106 o Insulators 5 x 107 2 x 103 1 x 195 9 x 106 (Organic)

Solenoid. Reactor 91dg.

Valves' Above 20' E1. 6 x 107 2 x 103 1 x 103 1 x 105 Pressure . Reactor Bldg.

Switches / Above 20' El.

Transmitters 2.1 x 10 8 2 x 103 1 x 103 1 x 105 Pump Motors: Reactor Bldg.

o Lube Above 20' El. 1 x 108 2 x 103 1 x 103 1 x 105

.o Insulation 5 x 107 2.x 103 1 x 103 1 x 105 o seals 5-x 106 2 x 103 1 x 103 2 x 105 Electronics: Reactor Bldg.

o Semiconductors Above 20' El. 1 x 108 2 x 10 3 1 x 103 1 x 105 o Capacitors 3 x 108 2 x 103 1 x 103 1 x 105 o Inductors 2 x 109 2 x 103 1 x 103 1 x 105 o Insulators 5 x 107 2 x 103 1 x 103 1 x 105 (Organic)

REFERENCES:

(1) "The Effects of Nuclear . Radiation on Electronic Components, Including Semiconductors," REIC Report No. 36.

Battelle Memorial Institute, Columbus, Ohio, October 1, 1964.

(2) " Space Fbterials Handbook," Second Edition, Document No. FE.-TDR-64-40, AF bbterials Laboratory, Wright-Patterson Air Force Base, Ohio (Prepared by Lockheed Missiles and Space Company, Sunnyvale, Calif.),

January 1965.

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