Text

To Environ Qualification of Electrical Equipment:Hb Robinson Eg Plant Unit 2, 90-day Rept
	ML19340D706
Person / Time
Site:	Robinson
Issue date:	11/01/1980
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML14174A745	List: ML14174A744; ML19340D706;
References
	IEB-79-01B, IEB-79-1B, NUDOCS 8101050194
	Download: ML19340D706 (100)
	v • d • e

.

O United States Nuclear Regulatory Commission Docket No. 50 - 261 License 20. DPR - 23 ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT H. 3. ROBINSON E. G. PLANT LWIT 2 NRC IE BULLETIN 79-01B (90-DAY REPORT)

CAROLINA POWER & LIGHT COMPANY RALEIGH, NORTH CAROLINA 4

FIRST ISSUE JUNE 1980 PREPARED BY: CAROLINA POWER & LIGHT COMPANY RALEIGH, NORTH CAROLINA Revision Date Revision Date 1.

1 8/21/80 4,

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2 11/1/80 5.

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ENFZRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT 57C Bulletin 79-013 (90-Day Report)

TABLE OF CONTENTS 1.

Table of Contents 11.

List of Figures 111.

List of Tables 1.0 General 1.1 Introduction 1.2 Preparation of Report 1.3 Report Parameters 2.0 MASTER LIST E ELECTRICAL EQUIPMENT REQUIRED E FUNCTION UNDER POSTULATED ACCIDENT CONDITION 2.1 Reference Sheet 3.0 ENVIRONMENTAL QUALIFICATION g ELECTRICAL EQUIPMENT REQUIRED E FUNCTION UNDER POSTULATED ACCIDENT CONDITIONS 3.1 Documentation Reference Sheet 3.2 Electrical Equipment Qualification Evaluation

4.0 CONCLUSION

S 5.0 REPORT QUALITY ASSURANCE R2 l

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APPENDICES Appendix A Calculations per Appendix 3 of IE Bulletin 79-013 to Determine Total Anticipated Radiation Appendix 3 Calculations per Appendix II to H.B. Robinson 10th Semi-i l

Annual Operating Report to Determine Submergence Depth R2 Appendix C Extracted Information Related to Radiation Exposure of l

Diallyl Phthalate l

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ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT NRC Bulletin 79-01B (90-Day Report)

LIST OF FIGURES l.3.1 H. B. Robinson Reactor Containment Radiation Level Measurement Locations

- 1.3.2 H. B. Robinson Containment Radiation Level Measurements 3.1.1 Containment Temperature vs. Time - LOCA 3.1.2 Conta1nment Pressure vs. Time - LOCA J

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ENVIRONMENT QUALIFICATION OF ELECTRICAL EQUIPMENT NRC Bulletin 79-01B (90-Day Report)

LIST OF TABLES 1.3.1 H. B. Robinson Calculated Radiation Accumulation 1.3.2 Reactor Coolant System Doses 1.3.3 Equipment Total Radiation Accumulation by Location and LOCA Operating Time 1.4.1 Equipment List for Safety Injection and Air Recirculation 1.4.2 Equipment List for Containment Isolation (Phase A) 1.4.3 Equipment List for Containment Spray Actuation and Containment Isolation (Phase B) 1.4.4 Equipment List for Long-term Accident Mitigation iii

1.0 GENERAL 1.1 Introduction The United States Nuclear Regulatory Commission Office of Inspection and Enforcement Bulletin 79-013 issued on Jan-uar, 17, 1980 required two responses--a 45-day and 90-day report. Carolina Power and Light responded to the 45-day report on March 10, 1980. The submitted volume is used as a base document for this 90-day report which will be referenced, extracted and updated within this volume to comply with the total require =ents of IE Bulletin 79-013 and subsequent NRC Region meeting minutes.

The 45-day report provided an overview listing of all elec-trical equipment within the Engineered Safety Systems which is required to funct1on under the postalated accident conditions and did not limit the listing to only Class IE equipment.

It also was concerned with equipment inside and outside the containment related to the detection of accident conditions, initial actuation of safety systems and the long-term miti-gation of postulated events.

The postulated events covered within containment are LOSS OF COOLANT ACCIDENT (LOCA) and MAIN STEAM HIGH ENERGY LINE BREAK (MSL3). Also covered was the HIGH ENERGY LINE BREAK (EEL 3) inside and outside of containment.

Review of the 45-day report indicates only a small number of electrical equipment is exposed to any actual harsh accident environment which would endanger functioning if not designed and qualified to withstand the postulated conditions. All other identified equipment must perform within near normal environments during and after postulated accident events.

Therefore this report will be limited to detailing in full the qualification of equipment identified as within the postulated accident environment. An area of exception is the RER pump l

compartment which will have high radiation level fluid cir-l culating through the pumps and piping during the mitigation aspect of accident condition. Therefore, electrical equipment i

outside of containment exposed to these radiation levels are also included in this report.

Additionally, the LOCA environment is more limiting when comparedwithSteam((neBreaksorHighEnergyLineBreaks

[

l within containment.

Therefore, the LOCA parameters will be used when qualifying or reviewing qualifications programs for l

all the accident conditions associated with the safety elec-trical equipment addressed in this report.

(1)Recent study performed for NRC IE Bulletin 80-04 (Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition) indicates a maximum containment pressure of 34.4 psig and a temperature of 257 F attained. Therefore, LOCA conditions still remain as limiting parameters for qualification.

. ~

1.2 Preparation of Report The preparation of this 90-day report proceeded as follows:

The A/E; Ebasco Services, Incorporated, was consulted to establish validity of data associated with original purchase orders and vendor preshipment testing.

The NSSS supplier, Westinghouse, was consulted to establish qualification coverage of electrical equipment in containment by WCAP or manufactured product testing programs.

Original manufacturers of containment electrical equipment and manufacturers of replacement equipment and hardware were contacted to provide qualification test programs / reports related to the types of equipment supplied.

Plant operating report data from original comaercial operation date n.s reviewed to determine any electrical equipment fai:

trends.

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Factors affecting operational life and accident condition performance were compiled and a program of preventative maintenance and/or replacement devised.

Reviewed the qualification parameters and compared them with current data to obtain realistic qualification values.

Compared current data against the previously submitted 45-day report and revised forms.

Compiled a testing program, where required, to establish or complete qualification of the safety electrical equipment where data is unavailable.

1.3 Report Parameters l

1.3.1 Flood Level 1

The H. B. Robinson containment lower level consists of a l

reactor vessel sump area and compartmented base floor.

The floor level elevation is at 228 feet.

The sump geometry will account for a filled volume of 68,000( )

i gallons of water. The containment geometry is such that each additional 120,000 gallons will add a one- (1) foot depth of R2 water within containment.

The anticipated volume of water available to flood the con-tainment during an accident is 451,000 gallons.

This is comprised of Refueling Water Storage Tank, Accumulators, Spray Addition Storage Tank, and Reactor Coolant Loop water volumes emptied within containment.

(2)See Appendix B to this report for calculations.

This will produce a floor flood level of approximately 3.2 feet or a flood elevation of 231.2 feet within containment.

Three (3) instruments mounted on the shield wall at a level of 230 ft. will be covered by the postulated flood elevation.

These are LT-459, LT-460 and LT-461, associated with pres-surizer water level indicating and alarming.

These instruments are not the only source of data for operator assessment and decision needed for HFT3 and LOCA situations.

E=ergency Instruction (E. I.1) states that information may be erroneous because of transmitter malfunction or failure due to accident environment in containment or abnormal conditions within the reactor coolant system.

Operators are instructed to use confir=ed data for pressurizer level and if level data is suspect, or lack?ng, to utilize other system information in decision situations.

The major contribution of pressurizer level information in an accident situation is to alert the operator that an accident has happened and initially aid in identifying the type of accident.

This occurs early in the accident scenerio. As pressurizer level is not essential infor=ation to achieve accident mitigation, their assumed failure under submergence will not necessitate relocation or replace =ent.

R2 The lowest mounted elevation of electrical equipment used for accident detection and mitigation is 231.5 feet.

Revision 1, dated August 21, 1980, of the 90-day report stated a flood level of 231.67 feet would be experienced and that certain instrumentation would be partially submerged. Only the nonelectronic sections of the instruments were affected and no operational difficulties were anticipated. With more exact data on flood level as reported in Appendix B, it was determined that the flood level would not achieve the 231.67-foot height but only 231.2 feet; therefore, the instruments previously listed as partially submerged are no longer in contact with flood water.

However, due to the close proximity of Instruments PT 444, PT-445 (Pressurizer Pressure Control), PT-455, PT-456, PT-457 (Pressurizer Pressure Safety Injection Signal), FT-474, FT-475 (Steam Line Break Monitor - Generator A), FT-484, FT-485 (Steam Line Break Monitor - Generator 3), FT-494 and FT-495 (Steam Line Break Monitor - Generator C) to the calculated flood level, additional radiation source exposure has been assigned and used to determine qualification (see Paragraph 1.3.2).

To establish additional distance between flood level and instruments, replacement transmitters for the above listed have been mounced with the =aximum height permissible in the instrument cabinets and still =aintain operability.

l For illustration purposes each identified class IE equipment location is listed and compared with the established flood i

level on the enclosed Environmental Oualification Equipment Recuired tc! Function Under Postulated Accident Conditions forms.

i

1.3.2 Radiat1on Inside containment accumulated radiation dosage for forty-(40)yearlifeandsingleaccidgntinc{gynt for H. B. Robinson had bean designated as 1.5 x 10 RADS.

This figure, when applied, was used for design performance and testing require-ment within eculpment specifications.

This figure is one of a series of calculated values associated with Westinghouse NSS supplied plants.

Initially a point kernel attentuation program modeled on the R. E. Ginna nuclear plant was used in 1971 to derive an accumulated dosage figure of),2.0 x 10 RADS. A refinement of this figure was performed byWestinghousetoaccommodatetherequgrementsofIEEE323, 1974. The resultant figure of 1.5 x 10 RADS in the original issuance of WCAP 8587, Westinghouse Environmental cualification of NSSS Class IE Eculpment, was stated as conservative and subject to revision when the source term issue was resolved.

Noted within this environmental program is the differential parameter for radiation exposure of equipment inside con-tainment by physical 1; cation.

Subsequent revision of WCAP 8587 accounts for location of equipment by level within containment to establish radiation exposure during both operation and under accident condition. A figure of 2.7 x 10' RADS is established as the most signi-ficant dosage accumulated.

Forconsiderationwithinth{sreport the radiation service condition of 1.4 x 10' RADS y, determined by use of IE Bulletin 79-013, Appendix B, will be used when reviewing in containment electrical equipment. This figure is represen-tative of H. B. Robinson parameters and depicts a thirty- (30) day integrated gamma dose.

Dose rates as listed in Table 1.3.1 have been used to evaluate equipment by location and application within contain=ent to determine forty- (40) year life dose accumulation. To support these radiation levels an evaluation was made of data accumu-lated during actual plant operation.

Six separate radiation readings at varied plant locations were collected during sequential years. Approximate locations are shown in Figure 1.3.1.

Radiation readings are charted on Figure 1.3.2 and Table 1.3.1 projects the data for a forty- (40) year operating period.

(3)FSAR Section 7A.

(4)See Appendix A to this report for calculations.

l Operating time after design event occurrence will determine the additional radiation dose accumulated.

After the appli-cation of a ten (10) percent margin, a total radiation dosage is listed in Table 1.3.3.

This total dosage will be used for comparison with equiprant tests performed and/or calculated values to determine everall qualification.

Outside of containment areas where recirculation fluids from inside containment are encountered the radiation dose of 6

4 x 10 RADS as stated within section 4.3.2 of IE Bulletin 79-013 represents the anticipated total dosane found a* *he outer diameter of pipe carrying Reactor Coolant water for a period of thirty (30) days after postulated event.

Use of Table 1.3.2 indicates a 4:1 ratio between pipe outer diameter and general area where affected electrical equipment-is locatgd. Therefore, a one- (1) month accumulated dose of l.0 x 10 LADS is what can be expected within the RHR com-partment when evaluating qualification of electrical equipment used to establish recirculation of containment sump water.

In containment following postulated accident, the flood level as established in Paragraph 1.3.1 will bring contaminated water close to and over some electrical transmitters. Transmitters (5) close to the water level-will be exposed to an additional radiation source. Source levels are established in Table D-8, Containment Sump Gamma Dose Rates and Integrated Doses Vs.

R2 Time, within NUREG 0588, Interim Staff Position og Environs 7utal Qualification of Safetv-Related Electrical Equipment, Ap adix D.

Since time will be required for the water level to rise close to the transmitters, the total integrated gamma dose at the surface utilized to calculate the total anticipated radiation exposure and listed in Table 1.3.3 is conservative.

Therefore, j

no additional radiation margin is required.

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(5) Transmitters affected are:

PT-444, PT-445, PT 455, PT-456, PT-457 - (30-minute operation required hour mininum used La calculation.)

FT-474,.FT-R%

i 475, FT-484, FT-485, FT-494, FT-495 day operation.

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H.3. ROBINSON REACTOR CONTAINMENT RADIATION LEVEL MEASUREMENT LOCATIONS

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TABLE 1.3.1 H. B. ROBINSON CALCULATED RADIATION ACCUMULATION m.

AREA (1)

YR. ACCUM.(2)

(2) 40 YR. ACCUM.

ELEV.(fc) l AI

1. CV Operating Deck (Pressure) 4.8 x 10

'l.9 x 102 280

2. CV Lower Level Polar Crane 5.7 x 10 2.3 x 10 233

3. CV Second Level-Seal fic;1e Rm.

8.5 x 10 3.4 x 10 254 5

4. Reactor Coolant Pump - Bay A 1.1 x 10' 4.4 x 10 243

5. Reactor Cnolant Pump - Bay B 2.8 x 10' l.1 x 10 243 6

3

6. Reactor Coolant Pump - 3ay C-9.6 x 10 3.9 x 10 243 3

7.2 x 10 2.9 x 10 (1) See figure 1.3.1 for locations.

(2) Calculations in (RADS)

(3) Total Containment (Averaged)

TABLE 1.3.2 REACTOR CCOLANT SYSTEM DOSES LOCATION DOSE r/hr PIPE CENTER 820 PIPE ID 470 PIPE OD 200 GENERAL AREA 30

TABLE 1.3.3 EQUIPMENT TOTAL RADIATION ACCUMLF1.ATION BT LOCATION AND LOCA OPERATING TIME I3)

Component Location level (ft) Time Cf Radiation Exp.

Acetdent Marsta Total Aac tcipated (Aceros.)

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FT-444 cv 231.5 30 MIN!'I 2.3 x 30 9.3 x 10' III 6

1.0 4 10 FT-445(2)

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3 I8I NO FT-456(2)

C7 231.3 30 MIN.(4) 2.3 x 10 9.5 x 10 3

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7 1.0 m to FT-457(2)

C7 231.5 30 MIN!'I 2.3 x 10 9.5 x 10 3

5 1.0 s 10 M-453 cv 231.5 30MDf'I 2.3 x 10 9.5 s 10 3

6 1.0 x 10 3

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6 LT-474 CV '

233 1 DAY 2.3 x 10 3.5 x 10 3.5 x 10 3,3,gg LT-475 C7 233 1 DAY 2.3 x 10 3.3 x 10 3.3 a 10 3.8 x 10 3

LT-476 C7 233 1 DAY 2.3 a 10 3.3 x 10 3.3 s 10 3.8 x 10 LT-477 C7 233 1 3AY 2.3 x 10 3.3 s 10 3.3 A 10 3.3 x 10 c-484 C7 233 1 3AY

3 x 10 3.3 x 10 3.3 x 10 3.4 a 10 0

6 LT-483 C7 233 1 DAY 2.3 x 10 3.3 s 10 3.3 x 10 3.3 x 10 3

0 LT-486 C7 233 1 DAY 2.3 x 10 3.3 x 10 3.5 s 10 3.C x 10 6

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LT-447 C7 233 1 DAY 2.3 x 10 3.3 a 10 3,3,gg 3,3, gg 3

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6 C -494 C7 133 1 DAY 2.3 a 10 3.5 x 10 3.5 x 10 3,g, gg 3

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6 LT-495 C7 233 1 DAY 2.3 x 10 3.3 x 10 3.5 a 10 3,3,gg 3

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6 LT-496 C7 233 13AY 2.3 x 10 3 5 a 10 3,3, gg 3.3 x 10 3

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6 LT-497 C7 133 1 3AY 2.3 a 10 3.5 x 10 3.3 x 10 3.8 x 10 LT-459 C7 230 30 MI't!N 2.3 a 10 9.5 x 10 0I 3

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5 9,3, gg LT-461 C7 230 30 ntw['I 2.3 x 10 9.5 x 10 9.3 a la' I3I 3

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0 M-474 C7 231.3 1 3AY 2.3 x 10 3.3 x 10 3.5 x 10 5.0 x 10 5.0 x 10' W 3

FT-475 C7 231.5 1 SAY 2.3 x 10 3.3 x 10' 3.3 a 10 6

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FT-484 C7 231.3 1 OAY 2.3 x 10 3.3 x 10 3.5 a 10 3,3,,, gg Q

6 FT-445 C7 231.3 1 DAY 2.3 x 10 3.3 x 10 3.3 x 10 3.0 x 10 5

0 FT-494 C7 23L.5 13AY 2.3 x 10 3.5 a 10 3.5 a 10 3.0 a 10 0

FT-492 C7 231.3 1 SAY 2.3 x 10 3.3 x 10 3.3 a 10 5.0 m 10 6(6) 5 6

FT-940 RAS,

230 30 DAYS 1.0 x 10 (6) 1.0 a 10 1.1 x 10 FT-943 2AB 230 30 OAYS 1.0 a 10 1.0 x 10 1.1 a 10 I

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FT-940 RAS 230 30 OAYS 1.0 x 10 1.0 x 10 1.1 s 10 6(0}

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FT-94 3 RAS 230 30 DAYS 1.0 x 10 1.0 a 10 1.1 x 10 291 3

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Sheet 1 of 2

TABLE 1.3.3 (Continued)

EQUIPMENT TOTAL RADIATION ACCUMULATION SY LOCATION

' AND LOCA OPERATING TDE I33

'is r g in Total Antictpated Component p.ocation Level (ft) Time of Itadiation Exp.

Accident I

(Aeprou.)

03cestion ft.0 vrs)(1) tadiation Exp.

(102) udtation Laooeure V-860A ItAS 212 30 DAYS 1.0 x to 1.0 x 10' 1.1 x 10' 1.0 x 10 1.0 x 10 1.1 x 10*

V-46CB RAS 212 30 DAYS 0

5 6

1.0 x 10 1.1 x 10 g,g, 33 V-861A RAB 212 30 QATS v 8615 RAS 212 30 DAYS 1.0 x 10 1.1 x 10' 1.1 x 10' 5

6 V-463A RA8 212 30 DAYS 1.0 x 10 1.1 x 10 g,g, gg T-4635 RA8 212 30 DAYS 1.0 x 10 1.1 x 10' 1.1 x 10 0

6 6

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CYC-381 RAB 240 30 DATS 1.0 x 10 1.1 x 10 1.1 x 10 Sct.cc0DS T12-7 C7 233

$ x13['I 2.3 x lis 9.3 x 10 3

9.3 x 10 3

5 5

712-9 C7 233 3 cr['I 2.3 x 10 9.5 x 10 9,3,gg 712-11 C7 233 5Mur['I 2.3 x 10 9.3 x to 3

5 9.3 x 10 712-13 C7 233 5 MIN ['

2.3 x 10 9.5 x 10 9.3 x 1E

'CTORS 575-1 C7 175 3 ERS.

1.9 x 10 3.1 x 10 3.1 x 10' 3.4 x 10 2

E75 2 C7 275 3 IGLS.

1.9 x 10 3.1 x 10' 3.1 x 10 3.6 x to 6

575-3 C7 275 3 BRS.

1.9 x 10 3.1 x 10 3.1 x 10 3.4 x 10 575-4 C7 273 3 ERS.

1.9 x 10 3.1 x 10 3.? a 10 3.4 x 10 ELECTRYCAL P TIONS Type 2 C7 234 -246 30 DAYS 2.3 x 10 1.4 x 10 7mPEurnE

%DCP*S 6

7 TE-'128 C7 243 (7) 1.1 x 10 1.3 x 10 7E-41D C7 243 (7) 1.1 x 10 i

1.3 x 10 l

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I TE-4223 C7 43 (7) 1.1 x 10 1.3 x 10 0

7(5) l TE-423 C7 243 (7) 1.1 x 10 1.3 10 (3) 6 7

l TE-4323 C7 243 (7) 1.1 x 10 1.3 10 I

6 7(3) l TE-432D C7 243 (7) 1.1 x 10 1.3 x 10 l

(N Ertrapolated from plant data (See Table 1.3.1) i l

l (2) Equipment located in instrument cabinets.

I I3' Calculation based on IE Sulletin 79-015. Appendiz *a.

ClwtTS/CRAPRS. Proceducee for Evaluating Casm Radiation Service Conditions.

I'Icharts /Craf

per 1r Sulletin 79-013. AppendfA 8 allow calculation to a sinimum of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

, exposure. This figure is conservative--no ra. gin required.

Total Integrated Radiation for accident condition (30 daye) per IE sulletin 79-015. Appendiz 3.

CHARTS /C2AP!t$.

Calculation bened on Accident Radiation figure - 2 x 10 ItADS.

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Not required for D8C=-used only for outside containment *SL3 protection.

Includew.ulded 7.9 x 10'ItADS for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months total incesrated sneuna does at the surface of containment euse vnter ( Per Appendix 0. Table D-4. NUKEC 0538).

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Includee added 1.15 x 10 ItA05 for 1 day totst intearatrJ ganuma dose at the surface of containment eump w. ster ( Per Appendix 0. Table 3-4. NURIG 0%88).

Sheet 2 of 2 8

1.3.3 Aging Since class IE electrical equipment; specified, designed and

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bu!.it for H. 3. Robinson did not require continued thermal and radiation aging called for under present qualification pro-grams each component has to be reviewed using broad spectrum data sources.

Three (3) sources have been selected as best meeting the requirements of IE Bulletin 79-OlB. These are:

(1) identification of similar equipment tested more recent than H. B. Robinson's equipment, (2) reinterpretation of existing test data performed during qualification testing but not specifically for aging purposes, and (3) evaluation of equipment materials for susceptibility to degradation due to thermal and radiation exposure.

Aging data available per the three categories above for listed components in section 3.0 of this report will be shown within the ENVIRONMENT, Qualification column of the System Component Evaluation Work Sheet forms. For the third category listed above degradation susceptibility sources utilized are:

(1)

NRC IE Bulletin 79-013, Environmental Oualification of, Class jgiEquipment, Appendix C, (2) Radiation Effects Design Handbook, Section 3, NASA CR-1787, and (3) A Review of Equipment Aging Theory and Technology, (Draf t Copy) EPRI RP-890-1.

Empirical data to date for H. B. Robinson gives a time base of ten (10) years' life for the electrical equipment identified within the Master List of Electrical Equipment Required to Function Under Postulated Accident Condition.

No significant f ailure_ rate has been experienced at H. B. Robinson with the listed equipment and only routine maintenance and alignment /

calibration procedures have been required.

1.4 Engineered Safety Feature Systems Electrical Equipment The following engineered safety feature systems were iden-tified as having electrical equipment required to function under the defined accident conditions:

o Safety Injection o Containment Isolation o Air Recirculation o Containment Spray d

f Electrical equipment associated with these systems is listed in Tables 1.4.1, 1.4.2, and 1.4.3.

The FSAR lists the above as Engineered Safety Feature Systems.

They are segregated by plant design systems and identified as to their functions by

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s.

o

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I use of reference sheet 2.1 within Section 2.0, Master List of Electrical Equipment Required tjl Function Under Postulated Accident Conditions of the 45-day report previously submitted.

The reference sheet 2.1 of this report segregates the equip-ment to be further evaluated and details additionally iden-tified systems and equipment.

Graphic portrayal of the listed instrumentation by accident function is located in figures 2.1.1 through 2.1.4 within the 45-day report submitted by CPSL in March 1980 and are not repeated for this submittal.

The environmental test profiles used for qualification pro-grams included in the 45-day report have not been resubmitted for this report.

It is noted that the formal documents referred and used to substantiate.,aalification contain the formated environmental profiles required. When these docu-ments are reviewed for qualification confirmation, the pro-4 files can be checked for plant accident parametric coverage.

4 i

l 4

5

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e.

Sheet 1 of 1 T.GLE 1.4.1 EQUIPM ST LIST FOR SAFETY INJECTION AND AIR RECIRCULATION 1.

V478, 488, 498 Main Feedwater Valve 2.

V479, 489, 499 Bypass Feedwater Valve 3.

V867 A, B 3oron Injection Inlet 4.

V870 A, B Boron Injection Discharge 5.

V866 A, B Hot Leg Injection (a) 6.

V744 A, B Core Deluge 7.

Safety Injection Pump Motor A, B, C

3.

Residual Heat Removal Pump Motor A, B 9.

Service Water Pump Motor A, B, C, D 10.

Service Water Booster Pump Motor A, B

11.

Containment Fans HVH-1, 2, 3, 4

12. Auxiliary Feedwater Pump Motor A, B 13.

Coctainment Spray Pump Motor A, B 14.

V878 A, B Safety Injection Pump Crosstie

Equipment used for Air Recirculation (a) Removed from automatic activation by Safety Injection Signal.

Sheet 1 of 2 TABLE 1.4. 2 EQUIPMENT LIST FOR CONTAINMENT ISOLATION (PFASE A) 1.

CVC-200A Letdown orifice isolation 2.

CVC-200B Letdown orifice isolation 3.

CVC-200C Lecdown orifice isolation 4.

CVC-204A Letdown line isolation 5.

CVC-204B Letdown line isolation 6.

PS-956A Sample line isolation (pressurizer steam) 7.

PS-956B Sample line isolation (pressurizer steam) 8.

PS-956C Sample line isolation (pressurizer liquid) 9.

PS-956D Sample line isolation (pressurizer liquid) 10.

PS-956E Sample line isolation (hot leg) 11.

PS-9567 Sample line isolation (hot leg) 12.

PS-956G Sample line isolation (accumulator) 13.

PS-956H Sample line isolation (accumulator) 14.

RC-HC-516 Pressurizer relief tank to gas analyzer 15.

RC-HC-519A Primary water to pressurizer relief tank 16.

RC-HC-519B Prtsary water to pressurizer relief tank 17.

RC-HC-553 Pressurizer relief tank to gas analyzer 18.

CC-HC-739 Component cooling from excess letdown heat exchanger 19.

SI-855 Nitrogen supply for the accumulators 20.

WD-1721 Reactor coolant drain tank pump discharge 21.

WD-1722 Reactor coolant drain tank pump discharge 22.

WD-1723 Containment sump to waste holdup 2pak l

23.

WD-1728 Containment sump to water holdup tank 24.

WD-1786 Vent header from reactor coolant drain tank 25.

WD-1787 Vent header from reactor coolant drain tank 26.

WD-1789 Gas analyzer from reactor coolant drain tank 27.

WD-1794 Gas analyzer from reactor coolant drain tank 28.

SGB-FCV-1930A Steam generator A blowdown line 29.

SGB-FCV-1930B Steam generator A blowdown line 30.

SGB-FCV-1931A Steam generator 3 blowdown line l

t l

l i

L

Shast 2 of 2 TABLE 1.4. 2 (Continued) 31.

SGB-FCV-19313 Steam generator B blowdown line 32.

SGB-FCV-1932A Steam generator C blowdown line 33.

SG3-FCV-19323 Steam generator C blowdown line 34.

SGB-FCV-1933A Steam generator A sample line 35.

SGB-FCV-lS33B Steam generator A sample line 36.

SGB-FCV-1934A Steam generator 3 sample line

&3 37.

SG3-FCV-1935A Steam generator C sample line 38.

SG3-FCV-1935B Steam ganerator C sample line 39.

RM-1 Radiation monitoring pu=p outlet 40.

RM-2 Radiation monitoring pump inlet 41.

RM-3 Containment outlet 42.

RM-4 Containment inlet 43.

IVSW-PCV-1922A Isolation valve seal water system 44.

IVSW-PCV-19223 Isolation valve seal water system 45.

EVAC-V12-6 Containment ventilation isolation valve 46.

EVAC-V12-7 Containment ventilation isolation valve 47.

EVAC-V12-8 Containment ventilation isolation valve 48.

EVAC-V12-9 Containment ventilation isolation valve 49.

EVAC-V12-10 Containment ventilation isolation valve 50.

EVAC-V12-ll Containment ';2ntilation isolation valve 51.

EVAC-V12-12 Containment ventilation isolation valve 52.

EVAC-V12-13 Containment ventilation isolation valve 53.

V841A, B Soron Injection Tank Recirculation

A

TABLE 1.4 3 EQUIP E T LIST FOR CONTAINMENT SPRAY ACTUATION AND CONTAINMENT ISOLATION PHASE B 1.

Containment Spray Pump A, B 2.

V880 A, B, C. D - Containment Spray Discharge Valves 3.

V381 - Reactor Coolant Pump Seal Water Return Line 4.

V626 - Reactor Coolant Pump Thermal Barrier Cooling Water Return Line 3.

V735 - Reactor Coolant Pump Thermal 3arrier Cooling Water Return Line 6.

V716 A, B - Reactor Caolant Pump Cooling Water Inlet Line 7.

V730 - Reactor Coolant Pump Bearing Oil Cooler Cooling Water Return Line 8.

V1-3A, Vl-3B, V1-3C - Main Steam Isolation valves

Equipment used for Containment Isolation Phase B l

l

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,v-,

..---,.-,am..-

e.--

v.

- +, - - -

= - ~ _ - _ -.. -.

i' Sheet 1 of 1 4

TABLE 1.4.4 s

EQUIPMENT LIST FOR LONG TERM ACCIDENT MITIGATION 1.

Residual Heat Removal Pump Motor A, B 2.

V869 Hot Leg Injection 3.

V860 A, B C.V. Sump to RHR Suction 4.

V861 A, B C.V. Sump to RHR Suction i

5.

V863 A, B RHR Discharge to SI/ Spray Suction f

A 1

)

A e

i 4

s i

b 4

i F

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..,... ~. -. _ _. _.. _.... _ _ _ - _ _. _.. _.. - _ -

ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT NRC IE Bulletin 79-01B (90-Day Report) 2.0 MASTER LIST OF ELECTRICAL EQUJPMENT REQUIRED l

TO FUNCTION UNDER POSTULATED ACCIDENT CONDITIONS a

j h

1 i

9 i

.j 1

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remy.-w,,,-,,,-y.m.--,,.puy, iq.w......m.wo---.,,,r,%,

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2.0 MASTER LIST OF ELECTRICAL EQUIPMENT REQUIRED TO FUNCTION UNDER POSTULATED ACCIDENT CONDITIONS The master list was developed by reviewing the FSAR, Westinghouse Technical Descriptions, Emergency Instructions, System Flow Diagrams and Plant Modifications.

Both safety-related equipment and associated equipment for accident mitigation were addressed and evaluated per the environmental conditions existing during their accident and post-accident application.

Systems presented on the Master List sheets represent the plant systems as shown on the flow diagrams contained in th FSAR and Westinghouse FWR NSSS drawings. To aid in cross referencing the plant syste=s to the Appendix A, Typical Equipment / Functions Needed for Mitigation of a LOCA or MSLB Accident, listing the following descriptive paragraphs apply.

In some cases equipment in the Plant System will appear in more than one Appendix A listing.

Engineered Safeguards Actuation All devices required for engineered safeguards initiation that are subject to a harsh environment are included in the Master List under the Safety Injection System. The processors of the signals '. logic cabinets) are not included as they are located in the control room area and are not subject to any accident harsh environment conditions.

R1 Reactor Protection Reactor Coolant System and Reactor Protection System list the instrumentation (RIDS, Pressure and Level Transmitters) required to provide Reactor Protection.

Containment Isolation Containment isolation valves are located outside centainment and close shortly after the accident occurs. They are not exposed to the containment harsh accident environment.

There-fore, they are not listed or evaluated in this section. As they are part of the Engineered Safety Feature Systems actuated by Phase A containment isolation signal, they are listed in Table 1.4.2, Equipment List for Containment Isolation (Phase A).

Main and Auxiliary Sta.am Line Isolation The main steam line isolation and break protection electrical equipment is included in the Master List under Main Steam System.

The electrical equipment for this system is located outside of containment and does not see the harsh accident environment and is not evaluated in this report.

Evaluation data on these ite=s is found in the H. B. Robinson 45-day Report.

Main and Auxiliary Feedwater Isolation Electrical Equipment related to LOCA and Steam Line Break for this title are listed under Feedwater System. The electrical equipment for this system located outside of containment does not see the harsh accident environment and are not evaluated in this report. Evaluation data on these items is found in the H. 3. Robinson 45-day Report.

Emergency Power The Emergency Diesel Generators, 4 kV Switchgear, 600 volt Load Centers and D.C. Systems are not part of this report as they are physically located away from containment and any other accident environment areas.

Containment Heat Removal HVAC System equipment list provides the electrical equipment provided for this function. This includes fan motors and con-tainment isolation valves.

Containment Ventilation The equipment required for this function is described under R1 Containment Heat Removal and Containment Isolation paragraphs of this section.

Control Room Habitability Systems and Vencilation for Areas Containing Safety-Related Equipment This equipment is not subject to any harsh environment condi-tions due to accident and is not included in this report for evaluation.

Service Water The equipment list for Service and Cooling Water System includes this title. As this equipment is not in the harsh accident environment, evaluation is presented in the 45-day Report.

Emergency Shutdown This function has been covered by reviewing the Plant Emergency Procedures and assuring that all electrical equipment stated pertaining to shutdown has been covered by listing and evaluation.

Post-Accidert Sampling and Monitoring 1

Equipment included for post-accident monitoring purposes is located in Main Steam System and Feedwater System listings.

Evaluation sheets are included in the 45-day Report.

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4 Radiation Monitoring The H. B. Robinson radiation monitoring system is not safety related and is not included in this report.

Safety-Related Display Information All of the display instrumentation associated with accident evaluation and accident mitigation is located in the control room or other non-harsh environment locations. Therefore, they are not included in the Master List.

1 4

I I

i ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT NRC IE Eulletin 79-01B (90-Day Report) 2.1 REFERENCE SHEET (2) Component is not exposed to DBE. No qualification required. Evaluation Work Sheet is not included in this report. See H.B. Robinson 45-day report on IE Bulletin 79-01B for data and evaluation.

(3) Component is not exposed to DBE but used for long term accident mitigation. Evaluation Work Sheet included in the report.

(4) Component was not included in H.B. Robinson 45-day 3

report on IE Bulletin 79-01B. Evaluation Work Sheet included in this report.

(5) Component 13 required for Main Steam Line Break detection.only. Evaluation Work Sheet included in this report.

(6) Component is being replaced by another nore significantly qualified component. See Evaluation Work Sheets this report for replacements. For known qualification information on this component seeH.3. Robinson 45-day report on IE Bulletin 79-013.

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l SYSTEM: SAFETY INJECTION COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment FT-940 (3) (4)

FLOW TRANSMITTER X

FT-943 (3) (4)

FLOW TRANSMInt.x X

PT-934 (3) (4)

PRESSURE TRANSMITTER X

PT-940 (3) (4)

PRESSURE TRAUSHITTER X

PT-943 (3) (4)

PRESSURE TRMSMITTER X

PT-950 (2)

PRESSURE TRANSMITTER X

PT-951 (2)

PRESSURE TRXNSMITTER X

l PT-952 (2)

PRESSURE TRANSMITTER X

l l

PT-953 (2)

PRESSURE TRANSMITTER X

PT-954 (2)

PRESSURE TRANSMITTER X

PT-955 (2)

PRESSURE TRANSMITTER X

l

~

LS-1925A (4)

LEVEL SWITCH X

L S-1925B (4)

LEVEL SWITCH X

~

l l

(1) When a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

Sheet 1 3

of

SYSTDt:

SAFEU INJECTION (continued)

COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Centainment Containment V-841A (2T VALVE, SOLENOID X

V-8413 (2)

VALVE, SOLENOID X

V-866A VALVE, MOTOR OPERATOR X

V-866B VALVE, MOTOR OPERATOR X

V-867A (2)

VALVE, MOTOR OPERATOR X

V-867B (2)

VALVE, MOTOR OPERATOR X

v-869 (3) (4)

VALVE, MOTOR OPERATOR X

V-870A (2)

VALVE, MOTOR OPERATOR X

l l

V-870B (2)

VALVE, MOTOR OPERATOR X

J l

V-878A (2)

VALVE MOTOR OPERATOR X

V 278B (2)

VALVE, MOTOR OPERATOR X

V-880A (2)

VALVE, MOTOR OPERATOR X

V-880B (2)

VALVE, MOTOR OPERATOR X

V-880C (2)

VALVE,. MOTOR OPERATOR X

(1) When a component is not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 2 3

of b _...., -.., -,

. - -.,, - ~.,, - - - -.,-,,.

n,

,..,,,-n

.,-..,,n--

,-v,.ng-

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,n..,

SYSTEM: SAFETY INJECTION (continued)

'I COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment V-880D (2)

VALVE, MOTOR OPERATOR.

X SI-A (2)

SAFETY INJECTION m ~n. vorn, X

SI-B (2)

SAFETY INJECTION PUMP, MOTOR X

SAFETY INJECTON l

SI-C (2)

PUMP, MOTOR X

Ch-A (2)

CONTAINMENT SPRAY PWD. MOTOR X

CS-B (2)

CONTAINMENT SPRAY MNP. VOTOR X

i

\\

1 1

1 (1) *Aen a component is not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

1 Sheet 3 3

of

l SYSTEM:

SAFETY INJECTION EQUIPMENT / COMPONENTS COMPONENTS Location Plant Identification Number (1)

Ceneric Name Inside Primary Outside Primary Containment Containment 2/C SHIELDED #16 INSTRUMENTATION CABLE X

X AMP #16/9 INSULATED TERMINAL LUG X

X 3/C #19/22 CABLE X

X EEAT SHRINK TU31NG CABLE SPLICE X

X C-3 ELECTRICAL PENETRATION X

D-2 ELECTRICAL PENETRATION X

D-8 ELECTRICAL PENETRATION X

D-9 ELECTRICAL PENETRATION X

SILICONE RUBBER TAPE #70 CONNECTION PROTECTION X

2/C 116, 3/C #16

0NTROL CABLE X

X 1C 500 MCM POWER CABLE X

(1) When a component i= not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

Sheet 1 of 1

SYSTEM:

REACTOR COOLANT t

I COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment LT--459 (6)

LEVEL TRANSMITTER X

LT-460 (6)

LEVEL TRANSMIITER X

LT-461 (6)

LEVEL "JRANSMITTER X

PT-444 (6)

PRESSURE TRANSMITTER X

' ~ ~ ~ ~

PT-445 '(6) ' '

PRESSURE TRANSMITTER X

~

'-~

PT-455 (6)

PRESSURE TRANSMITTER X

PT-456 (6)

PRESSURE TRANSMITTER X

PT-457 (6)

PRESSURE TRANSMITTER X

i

\\

l ~

(.1)'When a component is not identified by plant identification number, the l

=anufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 1 of 1 i

,-a

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^

SYSTEM:

REACTOR COOLANT EQUIPMENT / COMPONENTS i

COMPONENTS Location Plant Identification i

Number (1)

Generic Name Inside Primary Outside Primary Containment Containment l

2/C SHIELDED #16 INSTRUMENTATION CABLE X

AMP #16 INSULATED TERMINAL LUG X

su B-2 ELECTRICAL PENETRATION X

B-5 ELECTRICAL PENETRATION X

B-9 ELECTRICAL PENETRATION X

~'

CROUSE HINDS RPC-317-160-50IN/S08N CONNECTOR, ELECTRICAL X

v- " ~ '~~ ~

CROUSE HINDS RPC-117-150-POIN/P08N CONNECTOR, ELECTRICAL X

X

~.

l l

\\

(1) Wen a component is not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

Sheet 1 of 1

,,n

1 l

SYSTF21: MAIN STEAM COMPONENTS Location Plant Identification j

Number (1)

Generic Name Inside Primary Outside Primary

~

Containment Containment FT-474 (6 )

FLOW TRANSMITTER X

FT-475 (6 )

FLOW TRANSMITTER X

FT-484 (6 )

FLOW TRANSMITTER X

~

FT-485 (6)

FLOW TRANSMITTER X

FT-494 (6 )

FLOW TRANSMITTER X

FT-495 (6)

FLOW TRANSMITTER X

PT-474 (2)

PRESSURE TRANSMITTER X

PT-475 (2)

PRESSURE TRANSMITTER X

l PT-476 (2)

PRESSURE TRANSMITTER X

PT-484 (2)

PRESSURE TRANSMITTER X

PT-485 (2 )

PRESSURE TRANSMITTER X

PT-486 (2 )

PRESSURE TRANSMITTER X

PT-494 (2)

PRESSURE TRANSMITTER X

PT-495 (2)

PRESSURE TRANSMITTER X

PT-496 (2)

PRESSURE TRANSMITTER X

I (1) Vnen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 1

2 of

1 i

SYSTEM:

MAIN STEAM (Continued)

COMPONENTS Location Plant Identification l

Nu=cer (1)

Generic Name Inside Primary Outside Primary Containment Cont 2inment PT 464 (2)

PRESSURE TRANSMITTER X

PT 466 (2)

PRESSURE TRANSMITTER X

PT 468 (2)

PRESSURE TRANSMITTER X

Vl-3A (2 )

VALVE, SOLENOID X

71-30 (2)

VALVE, SOLENOID X

V1-3C (2)

VALVE, SOLEN 0ID X

l

-. ~.

l l

l

(.1) *n' hen a component is not identified by plant identification nu=ber, the

=anufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 2 og 2

SYSTEM: MAIN STEJM EQUIPMENT / COMPONENTS COMPONENTS Location Plant Identification Number (1)

Ceneric Name Inside Primary Outside ?rimary Containment Containment 2/C SHIELDED #16 INSTRL%UTATION CABLE X

X AMP vle INSULATED TERMINAL LUG X

X B-1 ELECTRICAL PENETRATION X

C-1 ELECTRICAL PENETRATION X

HEAT SHRINK TUBING CABLE SPLICE X

SILICON RUBBER TAFE 7d CONNECTION PROTECTION X

3/C #16 2/C #16 CONTROL CABLE X

CROUSE-HIliDS RPC-317-160-Soln /S0cN CONNECTOR, ELECTRICAL X

X RPC-ll7-150-POIN/P08N CONNECTOR, ELECTRICAL X

X i

i i

(1) 'Jhen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 1 of g

l

TDi

FEEDWATER COMPONENTS Location Plant Identifica tion Number (1)

'--._. ric Name Inside Primary Outside Primary Containment Containment LT-474 (6,

LEVEL TRANSMITTER X

LT-475 (6)

LEVEL TRANSMITTER X

LT-476 (6)

LEVEL TRANSMITTER X

LT-477 (6)

LEVEL TRANSMI m x X

LT-484 (6)

LEVEL TRANSMITTER X

LT-485 (6)

LEVEL TRANSMITTER X

LT-486 (6)

LEVEL TRANSMITTER X

LT-487 (6)

LEVEL TRANSMITTER X

LT-494 (6)

LEVEL TRANSMITTER X

LT-495 (6)

LEVEL 'JRANSMITTER X

LT-496 (6)

LEVEL TRANSMITTER X

LT-497 (6)

LEVEL TRANSMI " R X

i V-478 (2)

VALVE, SOLEN 0ID' X

V-479 (2)

VALVE, SOLEN 0ID X

(.1) 'a"nen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

(Rev-1) l Sheet 1 of 2

SYSTEM:

FEEDWATER (Continued)

COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment V-488 (2)

VALVE, SOLEN 0ID X

V-489 (2)

VALVE, SOLEN 0ID X

V-498 (2)

VALVE, SOLENOID X

V-499 (2)

VALVE, S0LENOID X

AFW-A (2)

FEEDWATER PUMP, MOTOR X

AFW-3 (2)

FEEDWATER PUMP, MOTOR X

i l

~

(1) When a component is not identified by plant identification nu=ber, the manufacturer, model number, serial number, etc., will be used.

l l

(Rev-1)

Sheet 2 og 2

~

SYSTEM:

FEEDtiATER EQUIPMENT / COMPONENTS COMPONENTS Location Plant Identification Number (1)

Ceneric Name Inside Primary Outside Primary Containment Containment 2/C SHIELDED #16 INSTRUMENTATION CABLE X

X AMP #16 INSULATED TERMINAL LUG X

X i

i HEAT SHRINK TUBING CABLE SPLICE X

C-1 ELECTRICAL PENETRATI'ON X

C-2 ELECTRICAL PENETRATION X

C-4 ELECTRICAL PENETRATION X

C-9 ELECTRICAL PENETRATION X

3/C #16, 2/C #16 CONTROL CABLE X

SILICON RUBBER TAPE CONNECTION PROTECTION X

CROUSE-HINDS RPC-317-160-S01N/S08N CONNECTOR, ELECTRICAL X

X CROUSE -HINDS RPC-ll7-150-PolN/?08N CONNECTOR, ELECTRICAL X

X 1/C 500 MCM POLTER CABLE X

[1) When a component is not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 1 og 1

SYSTEM:

AUXILIARY COOLING COMPONENTS Location Plant Identification Number (1)

Ceneric Name Inside Primary Outside Primary Containment Containment V-626 (2)

VALVE, MOTOR OPERATOR X

V-716A (2)

VALVE, MOTOR OPERATOR X

V-7163 (2)

VALVE, MOTOR CPERATOR X

V-730 (2)

VALVE, MOTOR OPERATCR X

V-735 (2)

VALVE, MOTOR OPERATC2 X

V-744A VALVE, MOTOR OPERATOR X

V-7443 VALVE, MOTOR OPERATOR X

RESIDUAL HEAT REMOVAL RHR-A (3 )

PUMP, MOTOR X

RESIDUAL HEAT REMOVAL RER-3 (3 )

PUMP, MOTOR X

~

(1)~ Man a component is not identified by plant identification number, the manufacturer, =odel number, serial number, etc., will be used.

(Rev-1)

Sheet i of g

>-e

.,w SYSTEM:

AUXILARY COOLING (RESIDUAL HEAT RDiCVAL)

COMPONENTS i

Locatica Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment V-960A (3) (4)

VALVE, MOTOR OPERATOR X

V-8603 (3) (4) yrtyp. yo7ep oprp3 mop X-V-861A (3) (4)

VALVE, MOTOR OPERATOR X

V-8613 (3) (4)

VALVE, MOTOR OPERATOR X

7-8A3A (3) (4)

VALVE, MOTOR OPERATOR

~

~'X

~ ' ~ ~ ~ ~

V-863B (3) (4)

VALVE, MOTOR OPERATOR

- ~ ~ -

X l

l l

(1) *inen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

Sheet 2 of 2

- - - - - - - - - ~~ ~~~~'-^ ~ ~"- ~

i SYSTDt:

AUXILIARY COOLING EQUIPMENT / COMPONENTS COMPONENTS

.l Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containme:.t Containment I

D-2 ELECTRICAL PENETRATION X

i

.i 1/C 500 MCM POWER CABLE X

AMP vl6/9 INSULATED TERMINAL LUG I

X REAT SRRINK TUBING CABLE SPLICE X

X SILICON RUBBER TAPE 70 CONNECTION PROTECTION X

3/C #19/22 CONTROL CABLE X

X 2/C #16/3C #16 CONTROL CABLE I

X (1) When a component is not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

Sheet 1

1 of

---~,..-,-,r.-

,_....m---

,,,--.,,,.-,.----,,,~,,.y..,,.w-.,,..,yc..-

,,--.,.,,..,_-...,_~.--.m_~..m.-.

SYSTEM: REACTOR PROTECTION l

COMPONENTS Location Plant Identification t

i Number (1)

Generic Name Inside Primary Outside Primary Containment Containment TE-412B (5)

TEMPERATURE ELEMENT X

TE-412D (5)

TEMPERATURE ELEMENT X

TE-4223 (5)

TEMPERATURE ELEMEiT X

TE-422D (5)

TEMPERATURE ELEMDIT '

X TE-4323 (5)

TEMPERATURE ELEMENT X

TE-432D (5)

TEMPERATURE ELEMENT X

1

'I (1) '4 hen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 1 of 1

i SYSTDt:

REACTOR PROTECTICN EQUIPMENT / COMPONENTS a

COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment 4/C SHIELDED #16 INSTRLHENTATION CA3LE X

X AMP #16 TERMINAL LUG X

C-4 ELECTRICAL PENETRATION X

C-9 ELECTRICAL PENETRATION X

CROUSE-HINDS RPC-317-160-50lN/S08N CONNECTOR, ELECTRICAL X

X CROUSE-HINDS RPC 117-150-P01N/P08N CONNECTOR, ELECTRICAL X

X

{

i l

i (1) %~nen a component is not identified by plant identification nu=ber, the manufacturer, model number, serial number, etc., will be used.

1 (Rev-1)

Sheet _1 of 1 1

J SYSTEM: SERVICE AND COOLING WATER COMPONE'iTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Conesin=ent Containment i

3,g, (3)

SERVICE WATER PUMP MOTOR X

SW-B (2)

SERVICE WATER PUMP, MOTOR X

3.,. C (2)

SERVICE WATER PUMP, MOTOR X

SW-D (2)

E U CE 'a*A M M,'

MOTOR X

g,,g, (3)-

SERVICE WATER BOOSTER PUMP, MOTOR X

SWB-B (2)

^

PUMP, MOTOR X

b

~ ~

(1) When a component is not identified by plant identification nu=ber, the manufacturer, model number, serial number, etc., will be used.

(Rev-1)

Sheet 1 of 1

SYSTEM: SERVICE AND COOLING WATER EQUIPMENT / COMP 05 TITS i

COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Pri=ary outside Primary Containment Containment f

1/C 500 MCM POWER CABLE X

t RUBBER TAPE MOTOR CABLE SPLICE X

l l

l i

l l

(1) When a component is not identified by plant identification number, the

=anufacturer, model number, serial number, etc., will be used.

Sheet 1 o f._1 r.--,,,-

-, -, - ~ ~ ~.., _ _., - -.. -. ~ - -,,,

,,,--,.,,.-,._,,,,,,,-.e

....---,n-,,._-..,-,s,,,.,n.,-,..,

.,-..e.--.

,-..n

,--.._,,r-

=-.

l 1

I 1

SYSTEM:

CHEMICAL & VOLT lME CONTROL COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Contain=ent Containment CVC-200A (6)

VALVE, SOLENOID X

CVC-2003 (6)

VALVE, SOLENOID X

CVC-200C (6)

VALVE, SOLESOID X

CVC-381 (3)(4)

VALVE, Motor Operator X

i l

t l

l (1) Wen'a compor.edt is not identified by plant identification number, the manufacture *:, model number, serial number, etc., will be used.

Sheet 1

1 of

SYSTEM:

CHEMICAL & VOLUME CONTROL EQUIPMENT / COMPONENT COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment 2/C #16 CONTROL CABLE X

3/C #19/22 (2)

CABLE X

g 2/C #16, 3/C #16 CONTROL CABLE X

X SILICON RUBBER TAPE #70 MOTOR CABLE SPLICE X

~ ~

HEAT SHRINK CABLE SPLICE X

TUBING I

C-3 ELECTRICAL PENETRATION X

~

R D-9 ELECTRICAL PENETRAIION X

~

(1) '4 hen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

Sheet 1 of 1

- - - - - -.. -. - - - _.. ~ _... -.. -.. -....... -. - - ~. _ -. - _., - _ _ -.. - - -. - - - -... - - -... -.. -. _. -

W g

I SYSTDi:

HVAC COMPONENTS Location Plant Identification Number (1)

Generic Name Inside Primary Outside Primary Containment Containment V12-6 (2)

VALVE, SOLENOID X

V12-7 (6)

VALVE, 50LCI0ID X

V12-8 (2)

VALVE, SOLEN 0ID X

i V12-9 (6)

VALVE, SOLENOID X

V12-10 (2)

VALVE, SOLEN 0ID X

V12-ll (6)

VALVE, SOLENOID X

V12-12 (2)

VALVE, SOLENOID X

l V12-13 (5)

VALVE, SOLENOID X

hTH-1 FAN, MOTOR X

HVH-2 FAN, MOTOR X

l HVH-3 FAN, MOTOR X

HVH-4 FAN, MOTOR X

l (1) '4 hen a component is not identified by plant identification number, the manufacturer, model number, serial number, etc., will be used.

(Rev-1) l Sheet 1 of 1

.-_y.__

__-,_.,.e____.r--e--rwem-*-*-r-

+re**-v-=~

e----*rewe--*-mew

-t-w---

e*

- - ~ ~ ~ -

-*-*-+-----==w

,**erw+

-v--------~

SYSTEM: hTAC EQUIPMENT /CCMPONENT COMPONE:."IS Location Plant Identifica tion Nu=ber (1)

Generic Name Inside Primarf Outside Pri=ary Contain=ent Containment 3/C #16, 2/C #16 CONTROL CABLE X

X AMP #16 TER.MINAL LUG X

X C-3 ELECTRICAL PENETRATION X

i C-6 ELECTRICAL PENETRAfION X

C-3 ELECTRICAL PENE~~ACION X

3-1 ELECTRICAL PENETRATION X

D-3 ELECTRICAL PE. rim 1 TION X

D-5 ELECTRICAL PENETRATION X

EEAT S* DRINK Ti!3ING CA3LE SPLICE X

1/C 500 MCM Pok'ER CA3LE X

l SILICON RL*33ER CONNEC* ION PROTICTION X

TM E d~n MOTOR CA3LE SPLICE (1) klien a ec=ponent is not identified by plan: identification nu=ber, the

=anuf ae:urer, model nu=ber, serial nu=ber, e:c., vill be used.

l Sheet 1 of 1

!L.

ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT NRC IE Bulletin 79-013 (90-Day Report) 3.0 ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EOUIPMENT REQUIRED TO FUNCTION UNDER POSTULATED ACCIDENT CONDITIONS S

g

..-,,. - - --ev.-

T

+e'-e v t-y- - s w e

-'"-$e y

e w w

-w yw w-

~a-t

--=r=v' e t'i+ W ww ' * -* e-e w

--e-'~ty

--=9'-v'*Tv5-1-"*te*-*#-'"ws

+-* e 4

v

-ev^t

ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT NRC IE Bulletin 79-OlB (90-Day Report) 4 3.1 DOCUMElfrATION REFERENCE SHEET 1.

Specification CPL-R2-E3 - Containment Structure Electrical Penetrations 2.

Westinghouse Letter CPL-77-550 (Electrical Penetrations) 3.

Crouse Hinds Quality Control Inspection Reports (Electrical Penetrations) 4.

NPR Penetration - Steam Incident and Helium Leakage Tests-with attached Stress Analysis Report 5.

Ebasco Specification: CPL-R2-E13, Electrical cable, I&C 6.

Ebasco Specification: CPL-R2-E14, Electrical Cable, 4160v and 480v 7.

Ebasco Specification:

CPL-R2-E-1, Motor Operators for Valves 8.

Westinghouse Specification:

E-676258, Motor Operated Valves 9.

Westinghouse Specification:

E-676270, Control Valves 10.

Ebasco PO NY-435227 to McIntosh Equipment Corp. for Containment Level Switches L1.

Ebasco Specification CPL-R2-IN-7, Level Switches 12.

Westinghouse Specification 676410, Instruments, general, l

inside containment 13.

Crouse Hinds Connector Data, Electrical Penetrations 14.

WCAP - 7410-L Vol. I Environmental Testing of ESF Related Equipment References not used within the 90-Day Report.

l

3.1 DOCUMENTATION REFERENCE SHEET (continued) 15.

WCAP - 7410-L Vol. II Environmental Testing of ESF Related Equipment 16.

WCAP - 9003 Fan Cooler Motor Unit Test 17.

WCAP - 7744-L Environmental Testing of ESF Related Equipment 18.

WCAP - 7829-L Fan Cooler Motor Unit Test 19.

WCAP - 8587 Environmental Qualification of Westinghouse NSSS Class IE Equipment 20.

H. B. Robinson Modification and Setpoint Revision No. 212 MSL3 Transmitter Shielding 21.

Postulated Pipe Failure Analysis outside of Containment 22.

Rosemount Test Report 117415 Rev. B, Model 1152 Transmitter 23.

Rosenount Test Report 3788, Model 1153A Transmitter 24.

Rosemount Product Data Sheet 2256, Model 1151 Transmitter 25.

Rosemount Test Report 97215A, Model 1151 Transmitter 26.

Rosemount Test Report 127227 Rev. A, Model 1151 Transmitter 27.

ASCO Service Bulletin, Solenoid Valves

28. WCAP-7153 Investigation of Chemical Additions for Reactor Containment Sprays 29.

Vendor Drawing 5379-4093 Motor Terminal Lead 30.

Emergency Instructions (E.I. - 1) Incident Involving Reactor coolant System Depressurization 31.

FSAR, pg. 5.1.2-28, Electrical Penetrations 32.

FSAR, pg. 7.5-11, Enviroaaental Capability

3.1 DOCL' MENTATION REFERENCE SHEET (continued)

33.

FSAR, pg. 6.3-14 to 6.3-20, Fan Cooler Evaluation 34.

FSAR, pg. 6.2-14, Motor Design Criteria 35.

FSAR, pg. 6.2-31,32, Pump & Valve motor Criteria 36.

FSAR, pgs. 6.3-4, 6.3-10, Air Recirculation System Criteria 37.

FSAR, pg. 6.4-12, Containment Spray System Criteria 38.

FSAR, Section 7, Amendment 7A, Component Environmental Testing Program

39.

Standard Manufacturer's Testing Program to Meet Design Criteria 40.

FSAR, pg. 7.5-11, Operating Time Requirements 41.

Rosemount Report 37821, Model 1153 Transmitter 42.

Limitorque Test Report FP-3271 i

43.

Qualification Tests for a Modular Penetration, Report AB-11/

12/13 44.

RAYCHEM, Technical Report F-C4033-3. Tests of Raychem Thermofit Insulation Systems Under Simultaneous Exposure to Heat, Gamma Radiation, Steam and Chemical Spray 45.

AMP Test. Report 110-11002, Qualification Test Report O

on AMP - Radiation resistant PIDG Terminals 46.

Continental Wire & Cable Company, Technical Report F-C2935, Test of Electrical Cables Under Simulated Post-Accident l

Reactor Containment Service 47.

ASCO Test Report No. AQS21678/TR, Revision A, Qualification Tests of Solenoid Valves by Environmental Exposure to Elevated Temperature, Radiation, Wear Aging, Seismic Simulation, Vibration Endurance, Accident Radiation and LOCA Simulation 48.

WCAP - 9157 Environmental Qualification of Safety Related Class 1E Process Instrumentation O

r %..5m

,,_,,.__y,_

y

...,y

__,,._e

.,.,-,n.r.c.,

.p,

+ _ - _

-..s.

,y-

3.1 DOCUMENTATION REFERENCE SHEET (continued) 49.

KERITE COMPANY - Letter dated August 5, 1980 enclosures: LOCA QUALIFICATION OF KERITE 1000 VOLT

,al FR/FR CONTROL CABLE LOCA QUALIFICATION OF KERITE 1000 VOLT HTK/FR POWER CABLE 50.

KERITE COMPANY - Letter dated October 21, 1980 g2 in response to CP&L letter, CO-02726, dated October 13,1980 requesting qualification data on use of SCOTCH 70 Silicone Rubber Tape

.,3 s.,

sp.,

=+-

-e

----"F-W'

' ' " - ' " * * ' " ' ' ' " ' ~ ~ * * " ' '**' * ' ' ' '

Facility: 11. 11. Robinson #2 Sheet 1 of 25 SYSTEM COMPONENT EVAI.llATIOff UORK SilEET s

I)OCllllENTATI011 REFERENCl-

"^

ENVIRONMENT EQlilPHENT DESCRIPTION CATION OllTSTANI)I NG Pa rame t er SpecifI-Qua1if1-SpecIIi-Qua1if1-HETil0D IT10lS catlou catlou cation cation (4)

System: SAFETY IN. LECT 10N Operating 30 DAYS 2 IIRS.

38 17 SIHUllrAN-Time g)

E0llS TEST Plant ID No. FT-940

'*I Component:

AHill ENT 287 38 17 SIMUI. TAN-t40NE Fl.0W TRANSMITTER E0tlS TEST Manufacturer:

Pressure A'1110S.

75 38 17 SIMULTAN-(PSIA)

FISilER & PORTER EOUS TEST Hodel Number:

Relative AMBIEttr 100 38 17 S IMUl. TAN-10B2496PilBABBB llumidity

(%)

EOllS TEST Function:

Chemical

.NOT l

SAFETY IN.IECTION Spray REQU IR ED Accuracy:

Spec:

1/2%

0 Radiation 1.I x 10 2 x 10 (3) 17 SEQllENTIAL Service:

TEST (5)

R2 ilEADER Fl.0W (Ilot 1.eg) 1.oca t i on :

g gg ng REACTOR AUXII.IARY lli.DG.

j Flood I.evel Elev: (2)

NOT Above Flood 1.evel: Yes Submergence REQlllRED No (1) Transmitter not exposed to DilE - 1.ong-term mitigation radiation exposure only (2) Not involved in containment flood postulation l

(3) See Section 1.3.2 (4) See Section 3.2.2 for evaluation.

(5) Test performed after 1.0CA simulated environmental exposure R1 i

I Facility:

11. B. Robinson #2 Sheet 2 of 25 SYSTEM COMPONENT EVAI.UATIDM WORK SilEET s

"^

ENVIRONilEllT DOCUllEllTATIOtt REFEREllCE EQUIPMENT DESCRIPTI0ti CATI 0tl OUTSTAt3D I N(,,

Pa rame te r Specifl-Qualifi-Spcciti-Qualifi-METil0D ITEllS cation cation cation cation (4)

System: SAFSTY IN.IECTION Operating 30 DAYS 2 IIRS.

38 17 SIMULTAN-NONE Time EOUS TEST Plant ID No. FT-943(II i

Temperature Component:

(Oy)

AMBIENT 287 38 17 SIMULTAN-fl0NE FLOW TRANSMITTER EOUS TEST Manuf a c ture r Pressure MU ' AN~

AntOS.

75 38 17 NONE FISilER & PORTER (psia)

EOUS TEST Model Number:

Relative 10B2496PilllABilB AMBIENT 100 38 17 SIMULTAN-llumidity NONE

(%)

EOUS TEST Function:

Chemical NOT SAFETY INJECTION Spray REQUIRED Accuracy:

Spec:

Demon:

l Radiation 1.1 x 10 2 x 10 (3) 17 SEQUENTIAL 6~

8 NONE TEST (5) gy Service:

IIEADER Fl.0W (Cold Leg) d Location:

RidACTOR AUXILIARY Bill 1. DIN (

E'"E Flood Level Elev: (2) g Above Flood Level: Yes Submergence REQUIRED tio (1) Transmitter not exposed to DilE - Long-term mitigation radiation exposure only (2) Not involved in containment flood postulation (3) See Section 1.3.2 (4) See Section 3.2.2 for evaluation.

(5) Test performed after LOCA simulated environmental exposure ll A I

Facility:

11. B. Robinson #2 Sheet 3 of 25 SYSTFJi COMPONENT EVAL.UATION WORK SilEET s

^ ' '

~

ENVIRONMENT DOCUMENTATIOtl REFERENCE EQlllPMENT ' ESCRIPTIOil CATIOli OUTSTANDING Parameter Specif1-,Qua1if1-

,vec1f1-Qua11f1-a FIETil0D ITEllS cation cation cation cation (4)

System: SAFETY I N.IECT ION Operating 30 DAYS 2 IIRS.

38 17 SIMULTAN-Time NONE Plant ID No. PT-934 (1)

EOUS TEST Component:

AMilIENT 287 38 17 S IMulli'AN-PRESSilRE TRANSMITTER Ma nu f ac t u re r:

Pressure ATHOS.

75 38 17 SIMULTAN-A NONF.

FISilER & PORTElt EOUS TEST Mode 1 Ntunber:

HelaLive 50EPIO4111CXA llumidity AMIIENT 100 38 17 SIMULTAN-(%)

EOUS TEST Function:

Chemical

.NOT 150RON INJECTION Spray REQUIRED Accuracy:

Spec:

Demon:

g g

Radiation 1.1 x 10 2 x 10 (3) 17 SEQUENTIAL Service:

TEST (5)

Al TANK llEADER PRESSIIRE I.oca t i on :

REACTOR AUXII.I ARY liLDG.

Flood f.evel Elev: (2)

NOT Above Flood I.evel: Yes Submergence REQUIRED No (1) Transmitter not exposed to DliE - I.ong-term mitigation radiation exposure only (2) Not involved in contaltunent flood postulation (3) See Section 1.3.2 (4) See Section 3.2.2 for evaluation.

gi (5) Test performed after I.0CA simulated environmental expusure

i Facility:

II. B. Robinson #2 Sheet 4 of 25 SYSTEM COMPONENT EVALUATI0tl WORK SilEET 4

^

~

ENVIRONMENT DOCUMENTATION REFERENCE EQtilPMENT DESCRIPTION CATION OUTSTANi>ING Pa rame t e r Speciff-Qualifi-S.peciti-Qualifi-tlETil0D ITEMS cation cation cation cation (4)

System: SAFETY INJECTION Operating 30 DAYS 2 IIRS.

38 17 SIMULTAN-

,,, i m e NONE Plant ID No. PT-940(1)

EOUS TEST

"*E rature AMillENT 287 38 17 SIMULTAN-NONE Component:

(oF)

PRESSURE TRANSMITTER EOUS TEST Ma nuf ac t ure r:

Pressure A1NOS.

75 38 17 SIMULTAN-(PSIA)

NONE FISilER & PORTER EOUS TEST Model Number:

Relative AMilIENT 100 38 17 SIMULTAN-50EP1041 lhamidi n 110NE EOUS TEST Function:

Chemical

.NOT SAFETY INJECTION Spray REQUIRED Accuracy:

Spec:

Demon:

0 Radiation 1.1 x 10 2 x 10 (3) 17 aEQUENTIAL tiONE Service:

TEST (5)

R2 IIEADER PRESSURE (llot Location:

Leg)

A I"E E

REACTOR AllXILIARY BUTI.1)INi Flood Level Elev: (2)

NOT Above Flood I.evel: Yes Submergence REQUIRED No (1) Transmitter not exposed to DilE - Long-term mitigation radiation exposure only (2) Not involved in containunent flood postulation (3) See Section 1.3.2

'4)

See Section 3.2.2 for evaluation.

('(5)

AA Test performed after LOCA s.imulated environmental exposure i

I I

Facility:

II. 11. Robinson #2 Sheet 5 of 25 SYSTEM COtlPONENT EVALUAT!or' WORK SilEET s

ENVIRON!!ENT DOCllflEllTATIOli REFEREllCE N"A EQUIPflEllT DESCRIPTION CATI 0tl GUTSTANDitlG Parameter Specif1-Qualifi-Spccilt-Qualifi-HET110D ITI'MS carfon catlon cation cation (4)

System: SAFETY INJECTION Operating 30 DAYS 2 IIRS.

38 17 SIMULTAN-(1)

Time EOUS TEST NONE Plant ID Ho. PT-943 Temperature SIMULTAN-Component:

(op)

AMBIENT 287 38 17 NONE EOUS TEST PRESSURE TRANSMITTER i

Manufacturer:

Pressure ATMOS

75 38 17 FISilEik & PORTER (PSIA)

EOUS TEST NONE Model Ilumber:

Relative SIMULTAN2 50EP1041BCXA Ilumidity AMBIENT 100 38 17 EOUS TEST

(%)

=

Function:

Chemical NOT SAFETY INJECTION Spray REQUIRED Accuracy:

Spec:

Demon.

6 0

Radiation 1.1 x 10 2 x 10

("l).

17 SEQUEtfrIAL NONE Service:

TEST (5) lgy llEADER PRESSURE (Cold Locatlon:

Leg)

^E "E REACTOR AUXILTARY llUILDIN(

Flood I.evel Elev:

(2)

NOT Above Flood Level: Yes Submergence REQUIRED No 1

(1) Tr a mitter not exposed to DBE - I.ong-term mitigation radiation exposure only

,[

(2) tiot involved in containment flood postulation (3) See Section 1.3.2 (4) See Section 3.2.2 for evaluation.

(5) Test performed after LOCA simulated environmental exposure SA

?

5 t

h i

i I

Facility:

11. B. Robinson #2 3,,g g

-- gg " ~

SYSTEM COMPONENT EVAI.UATION WORK SilEET 6

0 "^

ENVIRON? LENT DOCUMENTATIOti REF(REllCl EQUIPflENT DESCRIPTION CATION OUTSTANDING Parameter Specifi-Qualifi-S,pec i f f -

Qualifi-METilOD ITl]lS cation cation cation catton (5)

System: SAFETY INJECTION Operating SIMULTAN-Time 1 IIR 7 UAYN 30 l4 EOUS TEST NONE Plant ID tio. V-866A (1) n Component: MOTOR OPERATOR Temicratur (2) 308 35, 38 14, 17 SIMULTAN-NONE EOUS TEST SIMULTAN-Manufacturer: LIMITORQUE P

e (3) 75 35 14, 17 EOUS TEST NONE Model Number: SMB-00 Relative 100 100 35 14, 17 SIMULTAN-llumidity NONE g)

EOUS TEST Function: Il0T LEC Chemical 311 B03 H 00 3 3 SIMULTAN-INJECTION Spray n cil Nacil 14 NONE D

a EOUS TEST Accuracy:

Spec:

Demon:

6 8

Rad 1ation 1.0 x 10 2 x 10 (4) 17 SEQUENTIAL NONE Service: MOTOR OPERATED TEST (6)

VALVE-SIS Location:

U Aging 40 YRS 17 SEQUENTIAL NONE CONTAINMENT 241 TEST (6)

Flood Level C1 e 231.2'

~

g2 Above Flood Level: Ves X Submergence NOT I

No REQUIRED NOTES:

(1) Same data this sheet applies to V-866B.

3 i

(2) See accident profile - Temperature - Figure 3.1-1.

(3) See accident profile - Pressure - Figure 3.1-2.

t (4) See Section 1.3.2.

(5) See Section 3.2.3 for evaluation.

l (6) Test performed prior to I.0CA simulated environmental exposure I

\\

Facility:

!!. H. Robinson #2 Sheet 7 of 25 SYSTEM COMPONENT EVALUATI0tl WORK SilEET s

QUA FI-ENVIRONMENT DOCll!!EllTATIOli HlWLRENCF EQUIPMENT DESCRIPTION CATIOtl OUTSTANDIIIG Pa rame te r Spec 1f1-Qua1if1-Spec 1t1-Qua1if1-

!!ETil0D ITl?tS cation cation cation cation (3)

J I"I' System: SAFETY I!!JECTION Ope ating (1) 7 DAYS 30 14 NONE gg T Plant ID No. V869

^"~

Component: MOTOR OPERATOR AMBIENT 308 35 14,17 NONE (o g,,

E ES SIMULTAN-Manuf acture r: LIMITORQllE Pressure ATMOS.

75 35 14,17 (PSIA)

EOUS TEST Model Number: SMB-00 Relative SIMULTAN-AMBIENT 100 35 14,17 Humidity EOUS TEST

(%)

Function:110T LEC INJEC-Chemical NOT 11 B0 SIMilLTAN-JEC pray U

N H 4,17 MUS U.ST NONE M

Accuracy:

Spec:

Demon:

6 8

SEQUENTIAI Radiation 1.1 x 10 2.0 x 10 (2) 17 NONE TEST (5)

Service: MOTOR OPERATED VALVE R2

^

Aging 40 YRS.

17 NONE REACTOR AUXILIARY BLDG.

TEST (5)

Flood I.evel Elev:

(4)

NOT Above Flood Level: Yes Submergence APPI.ICABLE No 1

(1) To be used intermittantly during mitigation of LOCA (2) See Section 1.3.2.

(3) See Section 3.2.3 for evaluation.

.(4) Not involved in containment flood postulation (5) Test performed prior to LOCA simulated environmental exposure U

i

\\

Facility:

11. B. Robinson #2 l

Sheet L of n SYSfEM CollP0llENT EVAI.UAT Oil WORK SilEET N"^'

EllVI R0!! MENT DOCUMENTATI0ll 1:lWEREtiCE EQUlPHENT DESCRIPTION CATI 0tl OUTSTAt3 DING Parameter Specifi-Qualifi-Specif1-Qualiti-HETil0D ITI2IS cation cation cation cation (4)

System: SAFETY INJECTION Operating

0NTINUOUS NONE (5)

Time Plant ID No. LS-1925A Tem rature Component: LEVEL SWITCll (2)

NONE (5)

!!anuf ac ture r: MADISON Pressure (3)

NONE (5)

(PSIA) flodel Number: 560,2 Relative llumidity

~

(%)

Function: CONTAINMENT SUMl

.11 B0 NONE 3 3 W

Chemical WATER LEVEL MEASUREMENT NaOH Spray Accuracy:

Spec: 1/ 2' I n-Demon: cremeni Radiation 1.4 x 10 NONE (5)

Servicc: DETECT WATER LEVEL. CllANGES O I' I IEN'l' 228 Aging NONE (5)

Flood Level Elev: 231.2' Above Flood Level: Yes Submergence NONE N2 No X (1) Same data this sheet applies to LS-19258 (2) See accident profile - Temperature - Figure 3.1-1 l

(3) See accident profie - Pressure - Figure 3.1-2 (4) See Section 3.2.7 for evaluation (5)

Function to be superceded by two channels of analog measurement equipment.

No qualification testing required.

t t

4

t Facility:

11. B. Robinson #2 Sheet 9 of n SYSTEll COMPONEljT EVALUATION WORK SilEET I

a ENVIRONMENT DOCllMENTATION REFERENCl-

"^

EQUIPMENT DESCRIPT10tl CATION OUTSTANDIllC Spe.ifi-Qua]{fi-Specill-Qualifi-c Pa rame te r METil0D ITEMS cation cation cation cation (5)

SIMULTAN-System: AUXILIARY COOLINL Operating 5 MIN 7 DAYS 40 14 EOUS TEST NONE Time Plant ID No y_7 g (y)

"* E(," " "

(2) 308 35, 38 14, 17 S IMUl/ FAN-NONE Component: MOTOR OPERATOl<

(

)

EOUS TEST Manufac ture r:

Pressure

" " I' ^" ~

(3) 75 35 14, 17 fl0!!E LIMITORQUE (PSIA)

EOUS TEST Model Number:

Relative SMB-3

^

Ilumid t ty 100 100 35 14, 17 N0flE E S ES

(%)

Function: 1:

TOR CORE Chemical ll h03 11 B03 3

3 SIMULTAN-Spray N 0li N ull 14 a

a EOUS TEST Accuracy:

Spec:

Demon:

5 8

SEQUEllTIAL Radiation 9.5 x 10 2 x 10 (4) 17 Service: MOTOR-OPERATED TEST (6)

NONE VALVE-SI S gg Location:

SEQUENTIAL CONTAINMENT Aging 40 YRS 17 TEST (6)

NONE 245' Flood Level Elev: 231.2' Above Flood Level: Yes X Submergence R 2, No NOTES:

(1) Same data this sheet applies to V-744B.

(2) See accident profile - Temperature - Figure 3.1-1.

j (3) See accident profile - Pressure - Figure 3.1-2.

(4) See Section 1.3.2.

(5) See Section 3.2.3 for evaluation.

(6) Test performed prior to LOCA simulated environmental exposure lgl i

i i

Facility: 11. 11. Robinson #2 SheetIJ)_ of 25 SYSTEM COMPONENT EVAI.UATION WORK SilEET

"^

ENVIRONMErlT DOCilflENTATION REFERENCE EQUIPMENT DESCRIPTIOli CATION OUTSTAt3D ItiL.

Parameter Specifi-Qualifi-Specifi-Qualifi-METN01)

ITEtiS cation cation ration cation (4)

System: AUXILI ARY C001.ING Opesating (2) 7 DAYS 30 14 SIMULTAN-NONE Time EOUS TEST Plant ID No. V860A I$f, "

1.

Non Component: MOTOR OPERA'll)R AMBIENT 308 35 14,17

(

llanufacture r: LIMITORQUE Pressure IMUI.W-NONE ATHOS.

75 35 14,17 (PSIA)

EOUS TEST Model Number: SHB-1 Relative ANBIENT 100 35 14,17 SIMULTAN-NONE Ilumidi ty EOUS TEST

(%)

Function: CV SUMP TO RilR Chemical

.NOT '

11 1M)3 3

SIMUlTAN-SUCTION Spray REQUIRED Na' Oli 14,17 N1 EOUS TEST Accuracy:

Spec:

Demon:

6 8

SEQUEllTIAL Radiation 1.1 x 10 2.'0 x 10 (3) 17 TEST (6)

NONE Seivice: MOTOR OPERATED VALVE Location:

SEQUENTIAL n

REACTOR AUXILIARY BLDG.

4 TEST W N0E Flood Level Elev:

(5)

NOT Above Flood Level: Yes Submergence APPLICABLE No (1) Same data this sheet applies to,V860B (2) To be used intermittantly during mitigation of LOCA.

(3) See Section 1.3.2.

(4) See Section 3.2.3 for evaluation.

'( 5) Not involved in containment flood postulation (6) Test performed prior to LOCA simulated environmental exposure NA 1

l

i 5

i i

Facility: 1 1. 11. Robinson #2 Sheet A of,n_

SYSTEM COMP 0tlENT EVALUATION WORK SiiEET I

s ENVIRON!IENT DOCilliEllTATION I:WFERENCI N"A '

EQUIPflENT DESCRIPTION CATION OUTSTANDING Pa rame te r Specift-Qualifi-b.peciti-Qualiti-FIETHOD

..l?tS carion cation cation cat iou (4)

System: AUXILIARY C001.ING Operating (2) 7 DAYS 30 14 SIMULTAN-NONE Plant ID No. V861A Component: MOTOR OPERATOR AMilIENT 308 35 14,17 SIMULTAN-Il0NE o

EOUS TEST 4

Pressure Ma nuf ac ture r: 1.IMITORQUE ATHOS.

75 35 14,17 SIMULTAN-NONE (PSIA)

EOUS TEST tiodel Number: SMB-1 Rela t ive' AMilIENT 100 35 14,17 SIMUI. TAN-NONE ilumidi ty

(%)

EOUS TEST Function: CV SUMP TO RilR Chemical NOT ll B0 SUCTION 3 3 SIMULTAN-Spray REQUIRED 4'

Na0li EOUS TEST NA Accuracy:

Spec:

1 Demon:

6 8

Radiation 1.1 x 10 2.0 x 10 (3) 17 SEQUENTIAL NONE Se rv ice: MOTOR OPERATED TEST (6)

VAI.VE R1 l 1.oca t ion :

SEQUENTI Al' Aging 40 YRS.

17 110N F' REACTOR AUXILIARY lil.DC.

TEST (6)

Flood I.evel hiev: (5)

OT Above Flood Level: Yes Submergenc APPLICA!!LE No (1) Same data this sheet applies to V861B (2) To be used intermittantly durin'g mitigation of LOCA.

(3) See Section 1.3.2.

(4) See Section 3.2.3 for evaluation.

(5) Not involved in containment flood postulation (6) Test performed prior to I.0CA simulated environmental exposure g7 I

e

\\

Facility:

11. H. Robinson #2 Sheet n of n SYSTEtt COMPONENT EVAI.UAT13" Wolu( SilEET 6

ENVIRONt!ENT DOCUMEtiTATIOli I:lri;RENCI-

^

~

EQUIPttENT DESCRIPTION CATION OUTSTANDING Parameter Specifi-Qualifi-Sptcifi-Qualifi-IIETl10D ITIO!S cation cation cation cation (4)

System: AUXILIARY COOLING Operating (2) 7 DAYS 30 14 S IMul. TAN-NONE Time EOUS TEST Plant ID No. V863A Temperature AMBIENT 308 35 14,17 SIMULTAN-NONE Component: MOTOR OPERATOR

( F)

EOUS TEST

!!anufacturer: LIMITORQUE Pressure ATHOS.

75 35 14.17 SIMULTAN-NONE (PSIA)

EOUS TEST Model Number: SMB-00 Relative AMBIENT 100 35 14,17 simul. TAN-NONE ilumidity (1)

EOUS TEST Function: RllR DISCllARGE Chemical

.NOT ll B0 3 3 TO SI SPRAY REQUIRED Spray Na0li

~

R1 SYST121 EOUS TEST Accuracy:

Spec:

" ' " ^

Radiation 1.1 x 10 2.0 x 10 (3) 17 NONE Service: MOTOR OPERATED

. TEST (6)

VALVE R1

. cation:

Aging 40 YRS.

17 NONE REACTOR AUXILIARY BLDG.

TS

)

Flood I.evel Elev:

(5) fl0T

\\bove Flood Level: Yes Submergence 3

gppnggggi,g No (1) Same data this sheet applies to.V863B (2) To be used intermittantly during mitigation of I.0CA.

(3) See Section 1.3.2.

(4) See Section 3.2.3 for evaluation.

+

(5) Not involved in containment flood postulation (6) Test performed prior to I.0CA simulated environmental exposure l R1 I

Fa m ity.

it. L.

.abli.-.. #2 Sheet I3 of 25 SYSTEM COMPONF.NT EVALUATION HORK SilEET I

6 N"^

~

ENVIRONMENT DOCllMEtll'ATIOli REFERErlCl EQUIPMENT DESCRIPTION CATION OUTSTAllDING 4

Pa rame te r Specif1-Qua1if1-h.pecifI-Qua1if1-

?!ETil0D I f f.51S cation cation cation cation (5) j System: AUXILIARY Operating CONTINUOUS C0tlTINUOUS 34, 35 (4)

(2) 1 C001.ING Time Plant ID No. gg g Component: 110 TOR, PUMP AllBI.

ItibE Manuf ac ture r:

Pressure 15 15 35, 19 (4)

(2)

WESTINC110USE (PSIA)

Model Number: 506UPZ Relative Function: CIRCULATE SUMP llumidity AMBIENT AMBIENT 35, 19 (4)

(2)

WATER & ilORATED REFilELING

(%)

WATER TO REACTOR COOLANT SYSTEM-POST LOCA Chemical NOT' NOT Spray REQUIRED REQUIRED Accuracy:

Spec:

Radiation 1.1 x 10 2.0 x 10 19 (3) 18 SEQUENTIAL.

NONE Service: RESIDUAL. IlEAT TEST REMOVAI. PUMP - SIS Location:

40 yrs.

18 SEQUENTIAL NONP.

AUXILIARY I!Ull. DING TEST Flood I.evel Elev: N/A NOT Above Flood Level: Yes Submergence APPL.lCAllLE No

+

l NOTES:

(1) Same data this sheet applies to RilR-11.

(2) Motor not exposed to D13E, no qualification testing needed.

(3) See Section 1.3.2 (4)

Information to be obtained f rom manufacturer.

(5)

See Section 3.2.8 for evaluation i

l

\\

FacilILy:

11. B. Robinson #2

. Sheet 14 of 25 SYSTEM C0tiPONENT EVALUATION WORK SilEET ENVIRotalENT DOCUNEllTATION liliERENCF fi~

EQUIPMENT DESCRIPTION

.g.

OUTSTANDING Specifi-Qualifi Speciti-Qualifi-METil0D ITillS cation cation cation cation System: REACTOR PROTECTIOl Operat ng 1 IIR.

2 WKS.

21 48 E

Plant ID No. TE-412B (1)

Ten.perature SIMULTAN-NONE Component:

op)

(2) 320 21 48 EOUS TEST (5)

TEMPERATURE ELEMENT l

SIMULTAN-Manufacturer: ROSEMOUNT Pressure (3) 81 21 48 (PSIA)

EOUS TEST (5)

Model Number: 176KF Relative SIMULTAN-NONE 1

1 8

llumidity EOUS TEST (5)

I

(%)

l Function: MAIN oiEAM Chemical 11 B03 4g SIMULTAN-NONE S' ray Naf0ll LINE BREAK MONITOR p

EOUS TEST (5)

Accuracy:

Spec:

Demon:

Radiation 7

g SEQUENTIAL NONE 1,5 x 10 1.0 x 10 (4) 48 T -REACTOR TEST (6)

(5) k!}IOLANY I.0h #1 SIS GENERATION

$1 Location:

SEQUENTIAL Aging 40 YRS. +

48 CONTAINMENT 243' TEST (6) 2 WKS.

Flood Level Elev: 231.2' NOT R 2.

Above Flood Level: Yes X Submergence APPLICABLE, No NOTES:

(1) Same data this sheet applies to TE-412D (2) See accident profile - Temperature - Figure 3.1-1 (3) See accident profile Pressure - Figure 3.1-2

! (4)

See Section 1.3.2 (5) Not required for DBE - used only for outsideicontainment Main Steam line Break protection (6) Test performed prior to I.OCA simulated environmental exposure g2 l

e i

t

Facility: 11. 11. Robinson #2 Sheet E of E SYSTEtt Cor-IPONENT EVAL.UATIO!1 WORK SilELT s

^ ' ' '

~

ENVIRONf!ENT DOCUMENTATI0!1 REFERENCF EQlllPHENT DESCRIPTION CATI 0tl OUTSTANDIllG Pa rame t e r Spec 1f1-Qua1Jfi-S pec i i. I-Qua1ii1-HETi!0D ITI2tS catlon cation cation cation System: REACTOR PROTECTION Operating 1 IIR.

2 UKS, 21 48 S IMUI. TAN-NONE Time EOUS TEST (5)

Plant ID No. TE-42215 (1)

Temperature Component:

(2) 320 21 48 simul. TAN-NONF.

TEllPERATilRE El.EMENT

}

(5)

I> r e_

re (3) 81 21 48 SIMilLTAN-NONF fla nu f ac ture r: ROSEM0 LINT EOUS TEST (5)

Hodel thimber:

17,6 KF Relative fl2 100 100 21 48 SIMULTAN-NONE Ilumidi ty

(%)

EOUS TEST (5)

Function: MAIN STEAM Chemical II 110 SIMULTAN-NONE 3 3 4g LINE liREAK MONITOR Spray NaOli EOUS TEST (5)

Accuracy:

Spec-Demon:

7 8

SEQUENTIAL' Radiation 1.5 x 10 1.0 x 10 (4) 48 NONE

-REACTOR N

N Service: AV (5)

COOLANT LOOP #2

Nocal g y TION 40 YRS. +

F SEQUENTIAL Agin8 2 UK. POST 48 NONE CONTAINMENT 243' ACCIDENT TES,i (6)

Flood Level Elev: 231.2' NOT fil-Above Flood Level: Yes X Submergence APPL.ICAllLE j

No l' NOTES :

i f

i(1) Same data this sheet applied to TE-422n d

l(2) See accident profile - Temperature - Figure 3.>1-1

!(3) See accident profile - Pressure - Figure 3.1-2 (4) See Section 1.3.2

. (5) Not required for DisE - only used for outside containment Main Steam Line Break protection l(6) Test performed prior to LOCA simulated environmental exposure M

i I

Facility: 11. 11. Itobinson #2 Sheet 16 of 25 SYSTEM COMPONENT EVAI.UATI0tl WORK SilEET i

6 ENVIRONMENT DOCilHENTATInti REFERENCl-

^ ' '

EQUIPMENT DESCRIPTION CATION OUTSTAND I N(.,

Pa rame ter S,pecif1-Qua1if1-J, pec i f1-Qua1if1-METitoD ITE!!S cation cation cation cation i

System: REACTOR PROTECTION Operating 1 IIR.

2 WKS.

21 48 E

Plant ID No. TE-432B (1)

Temperature (2) 320 21 48 SIMUI. TAN-NONE Componer.t:

F)

EOUS TEST (5)

TEMPERATURE ELEMEtiT

" " ' ^ " ~

Pressure (3) 81 21 48 ttinu f ac ture r: ROSEMOUNT (PSIA)

EOUS TEST (5)

Model Number: 176KF Relative IMU. AN-NONE M

100 100 21 48 llumidity EOUS TEST (5)

(%)

Function: MAIN STEAM-Chemical NJ0li 48 EOllS TEST (5)

! ! 11 0 SIMUL. TAN-ENE 3 3 LINE BREAK MONITOR Spray Accuracy:

Spec:

0 Radiation 1.5 x 10 1.0 x 10 (4) 48 SEQUENTIAL NONE Service: T -REACTOR (6)

W COOLANT LO NghATIbrY' l.oca t ion :

40 YRS. F SEQUENTIAL Aging CONTAINHENT 243' 2 WKS.

48 TEST (6)

Flood Level Elev: 231.2' NOT N

Above Flood Level: YesX Submergence APPLICAllLE No NOTES:

(1) Same data this sheet applies to TE-432D (2) See accident profile - Temperature - Figure 3.1-1 (3) See accident profile - Pressure - Figure 3.1-2 (4) See Section 1.3.2 (5) Not required for DilE - only used for outside containment main steam line break protection (6) Test performed prior to LOCA simulated environmental exposure lR2

r l

5

\\

Facility:

11. B. Robinson #2 Sheet n of g SYSTEtt COMPONENT EVAI UATION WORK SilEET s

" ^ ' '

~

ENVIRONMEt1T DOClillENTATION REIV RI:NCE EQUIPMENT DESCRIPTION CATION OUTSTANDING Pa rame te r Specifi-Qualifi-Specifi-Qualifi-METil0D ITEMS cation cation cation cation (5)

System:

IIVAC Operating 3 hrs.

24 hrs. +

36 16 None I""I'^""

Time Test Plant ID No.

IIVil-1 (1)

"*E$[,") "'"

Component:

MOTOR, FAN I""II""" ""

(

(2) 315 36 16 None Test Simultaneous Manuf acture r:

Pressure (3) 75-95 36 16 None WESTINGil0USE (p3IA)

Test Model Number:

685.5-S Relative Simultaneous llumidity 100 100 36 16 Tm

(%)

Function:

TRANSFER llEAT Chemical 11 B0 Il B0 Simuhaneous FROM CONTAINHENT TO 3 3 3 3 34 16 None Spray SERVICE WATER Nach Na0h Accuracy:

Spec:

Demon:

I Radiation 6

"9""""

3.4 x 10 1.41.x10 (4) 15 None Service:

CONTAINMENT Test (6)

FAN COOLER 4

y i

f-S"9"""'I"I Aging 40 yrs.

15 None i

CONTAINMENT 275' Test (6)

Flood I.evel Elev: 231.2' M

Above Flood Level: Yes X Submergence NOT l

{

No APPLICABI.E i

1 NOTES:

(1) Same data this sheet applies to llVil-2, ilVil-3, llVil-4 I

(2) See accident profile - Temperature - Figure 3.1-1 (3) See accident profile - Pressure - Figure 3.1-2 (4) See Section 1.3.2.

e (5) See Section 3.2.8 for evaluation.

]

(6) Test performed on selected motor components - not part of LOCA simulat;ed environmental exposure g

I i

8 1

f

\\

l

\\

t Facility:

11. B. Robinson #2 Sheet 18 of 25 SYSTEM COMPONENT EVALUATION UORK SilEET s

N "^ '

ENVIRONMENT DOCllMENTATION lilTI:RENCE EQUIPHENT DESCRIPTI0tl CATI 0tl OUTSTANDitlG Parameter Specif1-Qualifi-Specifi-Qualifi-HETil0D ITDIS cation cation cation cation System:

ALL Operating CONTINUOUS 105 hrs.

I 2,43 SIMULTANEOUS NONE Time TEST Plant ID No. SEE NOTE (1)

Component: El.ECTRICAL Temperature PENETRATION (OF)

(2) 340 1

2s'3.4,43 SIMULTANEOUS NONE TEST

=

Hanuf ac ture r:

Pressure (3)-

75 1

2,3,4,43 SIMULTANEOUS NONE CROUSE-IllNDS (PSIA)

TEST Hodel Number:

Relative 1.2.2 (745) 1.2.$ (751) liumidity 100 100 1

2,4,43 SIMULTANEOUS NONE 1.2.2 (747)

(%)

TEST 1.2.4 (749)

Function:

ACCIDENT Chemical U3 3 CONDITION MONITORING

MUI, OUS NONE

~

Spray Na0li TEST Accuracy:

Spec:

Radiation 1.4 x 10 2.13 x 10 (6) 43 SEQUENTIAL NONE (5)

Service: PROVIDE CA11LE TEST (7)

CONTINUITY TilROUGil

- g,gg[l[pgMENTSHELL 524 hrs. @

Aging 40 150 C SEQUENTIAL NONE g2 t

CONTAINMENT 234' - 246 4

(40 yrs:

TEST (7)

, Flood Level Elev: 231. 2'

^

NOT Ill

!Above Flood I.evel: Yes X Submergence APPLICABLE No NOTES:

(1) Data this sheet applies to penetrations 11-1,u-2,B-5,B-9,C-1,C-2,C-3,C-4,C-6,C-8,C-9,D-1,D-2,D-3,D-5,D-8,D-9 (2) See accident proffle - Temperature - Figure 3.1.1

(3) See accident profile - Pre.ssure - Figure 3.1.2

!(4) See Section 3.2.1 for evaluation

.(5) Qualification established for penetration cartrfdge only. Pigtail cab [e requires separate testing as t

reported in Section 3.2.1

,(6) See Section 1.3.2 (7) Test performed prior to.LOCA simulated environmental exposure NA 1,

f

I Facility:

11. B. Robinson #2 Sheet _19 of 25 SYSTEM COMPONENT EVALUATION WORK SilEET, s

" ^ ' '

~

ENVIRONMEllT DOCllMENTATION HEFERENCE EQUIPMENT DESCRIPTION CATION OUTSTANDING Pa rame te r Specifi-quallri-Speciti-Qualifi-MET 1101)

ITEMS cation cation cation cation (6)

System:

ALL Opc a Ing

,}

NONE L llR.-1 DAY 67 IIRS.

38 23 Plant ID No. SEE NOTE (1)

"*E Component:

SIMULTAN-( F)

(2) 350 38 23

""" h, THANSMITTER EOUS TEST SIMULTAN-Manu f acture r: ROSEMOUNT (3) 135 38 23 EOUS TEST Model Number:

1153A Relative llumidity 100 100 23 SIMULTAN-NONE

(%)

TEST Function:

Chemical ll M II 3 3 3 3 HEPLACEMENT COMPONENT

' Spray 23,41 X)

Na011 Na011 Accuracy:

Spec:

-+ (%

Demon:

Radiation 5 0 x 10 4.4x10 (S) 23 SEQUENTIAI (4)

Service:

82 NOT WITilIN

. Location:

Aging MFCR. TEST (4)

CONTAINMENT PROGRAM Flood Level Elev: 231.2'

~

lR2-

Above Flood Level: Yes Submergence i

No NdTES:

(1) Replacement ransmitter to be supplied for:

PT-444, PT-4/ 5, PT-4 55, PT-456, PT-457, l.T-474, LT-4 7 5, LT-476, LT-477, LT-484, I/r-486, LT-487, LT-494, LT-495 LT-496, LT-497, LT-459, LT-460, LT-461, FT-474, FT-475, FT-484, FT-485, FT-494, FT-495, LT-485

'(2) See accident profile - Temperature - Figure 3.1-1 (3) See accident profile - Pressure - Figure 3.1-2 (4) Replacement transmitters tested under IEEE 323-1971 format, Rosemount currently performing transmitter testing to meet IEEE-323-1974 requirements.

($) See Section 1.2.3 (7) Test performed prior to LOCA simulated environmental R2 i

! (M s....

s,.c e inn 1 9 i rnr

<m, s..u s,m

" nn u u ro

f Facility: 11. 11. Robinson #2 Shee t _21 o f _21 SYSTDI COMPONENT EVALUATION WORK SilEET ENVIRONMENT DOClH1ENTATION REFERENCI-

^ ' '

~

EQUIPMENT DESCRIPTION CATION OUTSTANDING Pa rame te r Speciff-Qualifi-Speci11-Qualif1-HET110D ITEMS cation cation carton cat ion (5)

System:

All.

Operating 5 min.

30 days 40 47 Time Test None Plant ID No. SEE NOTE (1)

"*Ef")"

(2) 346 40 47

""I*"""

Component:

(F Test None Sol.ENOID, val.VE Pressure I""IE"""

Manufacturer:

ASCO (3) 125 40 47 None (PSIA)

Test Model Number:

Relative NP831665E I"" ' "" "*

liumidity 100 100 40 47 None NP8316E35E Test

(%)

206-381-2U Function:

Chemical ll B0 "3

3 47 3 3 REPLAC&fENT COMPONENT Spray Na0li Na0ll Accuracy:

Spec:

Demon:

5 8

Sequential Radiation 9.5 x 10 2.0 x10 (4) 47 None Test Service:

(6)

, Location:

40 yra, t Sequential Y'"'

}

CONTAINMENT 283' Test (6)

None Flood Level Elev: 231.2' Not Above Flood Level: Yes X Submergence Applicable El i

No

' NOTES:

i l(1) Replacement solenoid valves to be supplied for:

V12-7, V12-9,'!*2-1;1, V12-13, CVC-200A, CVC-200B, CVC-200C lR2 (2) See accident profile - Temperature - Figure 3.1-1 4 (3) See accident profile - Pressure - Figure 3.1-2:

(4) See Sectlon 1.3.2 3

(5) See Sectlun 3.2.6 for evaluation

, (6) Test performed prior to LOCA simulated environmental exposure l

l42

i 4

Facility: 11. 11. Robinson #2 Sheet 21 og 25 SYSTEM COMPONENT EVALUATION WORK Stil:ET a

ENVIRONilENT "A

~

DOCutlENTATION REFERENCE EQUIPMENT DESCRIPTION CATIOtt OUTSTA!!DIllG Pa rame te r Specif1-Qualifi-Speciti-Qualifi-FIETi!0D ITillS cation cation cation cation System:

ALL Operating SIMULTANEOU:

Time CONTINUOUS 240 hru.

46 TEST Plant ID No.

'*Ef")"#"

Component:

CAllt.E (F

(2) 340 5

46 4/C #16, 2/C #16, TEST Shielded Manuf ac ture r:

Pressure (3) 115 46 SIMULTANEOUS NONE CONTINENTAL WIRE & CAllLE Model Number: CC2115 Relative 100 100 46 SIMULTANEOUS NONE llumidity (g)

TEST Function: FIELD CABLE Chemical H E0 Spray 3 3 46 SEQUENTIAL NONE Accuracy:

Spec:

TEST Radiation 1.4 x 10 l.0 x 10 (1) 46 SEQUENTIAL NONE

~

Service: INSTRUMENTATION TEST Location: CONTAINHENT Aging 5

(4) l

. Flood Level Elev: 231.2' R1

,Above Flood Level: Yes Submergence NOT po APPLICAllLE

! NOTES:

(1) See Section 1.3.2 l((3)

2) See accident proffle - Temperature - Figure 3.1.1 See accident profile - Pressure - Figure 3.1.2 (4) See Section 3.2.4 for evaluation i

h

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i Facility:

11. B. Robinson #2 Sheet 23 of 25 SYSTEM COMP 0llENT EVALUATIOtl UORK SilEET i

0 "^

~

EllVI RONilENT DOCilflEllTATI0tl itif t:REllCE EQUIPflENT DESCRIPT10tl CATI 0tl OUTSTANDIllG Parameter Specif1-Qua1if1-Spec 1f1-Qua1if1-tlET110D IT121S cation cation cation cat ion (3)

System: CllEMICAL & VOLUME Operating (1) 7 DAYS 30 14 NONE Time 3.l E T Plant ID No. CVC-381 Tem ture Component: MOTOR OPERATOR AMBIENT 308 35 14,17 110NE E S SIMULTAN-Manufacture r: LIMITORQUE Pressure ATHOS.

75 35 14,17 (PSIA)

EOUS TEST Model Number: SMB-00 Relative IMU AN-AMBIENT 100 35 14,17 NONE Ilumidity EOUS TEST

(%)

Function: REACTOR COOLANT Chemical

.NOT 11 B0 3

Spray REQUIREI)

NdON 14,17 SIMlIL PUMP SEAL WATER RETURN NONE y

Accuracy:

Spec:

Radiation 1.1 x 10 2.0 x 10 (2) 17 SEQUENTIAL NONE MOTOR OPERATED Service:

VALVE TEST (5) i Location:

240' AA REACTOR AUXILIARY BLnc.

Aging 40 YRS.

17 NONE TEST (5)

Flood Level Elev: (4)

NOT Above Flood Level: Yes Submergence APPLICABLE No (1) To be used intermittantly during mitigation of LOCA (2) See Section 1.3.2 4

(3) See Section 3.2.3 for evaluation (4) Not involved in er.ntainment flood postula? ion l (5) Test perfor. mend prior to LOCA almulated environmental expo sure 4

t I

3 I

j

l l'acility:

11. n. Robinson #2 Sheet 24 og 25 SYSTEM COMPO!1ENT EVAL.UATIOll WORK SilEET i

^ ' '

ENVIRONilENT I)0CilNEllTATlotl REFERENCI EQUIPflENT DESCRIPTION CATIOff OUTSTANDING Parameter Spectf1-Qua1if1-Spec 111-Qualif1-flETil0D ITI0lS cation catton cation catton i

System:

ALL Operating CONTINUOUS 30 days 44 Time TEST Plant ID No.

Component: CABl.E SPl. ICE

(

')

(2) 357 44 Manu f acture r: RAYCilEM Pressure (3) 85 44 SIMULTANEOUS NONE (PSIA)

TEST Hodel liumber:

Relative 1000-12N, 500-12N.

Ilumidi ty 100 100 44 SIMULTAllEOUL NONE 300-12N, 200-12N, (y,)

TEST 115-6N, 070-6N Function: SINGLE CONDUC-Chemical II B0 TOR AND MULTICONDUCTOR 3 3 44 simul.TANEOUS NONE Spray CAllLE SP Na0ll TEST Accu racy,J.IC INGbpec:

Radiation 1.4 x 10 2.0 x 10 (5) 44 SEQUENTIAL NONE Service:

ELECTRICAL S (6) y PI:NETRATIONS 7 days 0 Aging 302"F 44 SIMULTAllEOU!

NONE ICONTAINMENT 234' - 246' 5 x 10 RAD TEST Flooil I.evel Elev: 231. 2' NOT R1 Above Flood I.evel: Yes X Submergence APPLICAllLE No 4

NOTES:

(1) Plant procedure developed and approved for installation and checkout - M-521 (Revision 0)

(2) See accident profile - Temperature - Figure 3.1.1 (3) See accident profile - Pressure - Figure 3.1.2 (4) See Section 3.2.5 for evaluation j

(5)

See Section 1.3.2 7

g (6) Test performed prior to (5 x 10 R) and after (1.5 x 10 R) 1.0CA simulated environmental exposure l

t

i Facility:

11. B. Robinson #2 Sheet 25 of 25 SYSTEM C0!!PONENT EVAL.UAT10!! WORK SilEET l

6 ENVI RollllENT DOCllllEt1TATI0tl HEFERENCI-

^

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EQUIPilENT DESCRIPTION CATIO!!

OUTSTAtlDING ParameLer Specif1-Qua1if1-S,pec 1 f 1-Qua1if1-HETilOD IT F'IS cation cation cation cation System:

AI.I.

Operating SIMULTANEOUS NONE Time CONTINUOUS, 4 days 45 y

TEST Plant ID No.

Temperatur S IMUllfANEOUS N0flE Cong.onen t : TERMI NAl.S,

350 45 gy

(

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fEST CAllLE S

UI.

EOUS NONE

)gy llanu f acture r:

AMP Pressure (3) 137 45 (PSIA)

TEST tiodel Number: 53548-1 Relative (wire size S

U NONE llumidi ty 100 100 45

- 16).

(g)

TEST Function: CONDUCTOR Chemical 11 B0 SIMULTANEOUS NONE BUTT SPLICE 3 3 45 Spray

.g,EST Na0ll Accuracy:

Spec:

0 Radiation 1.4 x 10 2.0 x 10 (5) 45 SEQUENTIAL NONE Service: El.ECTRICAl.

TEST (6)

RJ PENETRATIONS LocaLlon:

CONTAINMENT 234' - 246 Aging (7)

, Flood Level Elev:231.2' gy

' Above Flood Level: Yes X Submergence PPLICABI.E No i

NOTES:

(1) Plant procedure developed and approved for installation and checkout - M-521 (Revision 0)

(2) See accident profile - Temperature - Figure 3.1.1

,1 (3) See accident profile - Pressure - Figure 3.1.2 (4) See Section 3.2.5 for evaluation (5) See Section 1.3.2 (6) Test performed prior to I.0CA simulat ed environmental exposure E2 (7)

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3.2 Electrical Eouipment Qualification Evaluation 3.2.1 Electrical Penetrations and Connectors The H. B. Robinson Nuclear Power Plant electrical penetrations are cartridge types with provisions for continuous pressuri-f zation. They were manufactured by Crouse-Hinds Company,.

(Syracuse, N. Y.) to a Westinghouse design and specification CPL-R2-E3.

Location within containment forms a grid pattern extending from elevation 234 feot to 246 feet.

This places the penetrations above the established containment flood level of 231.2 feet. The electrical penetrations utilized by iden-tified safety class electrical equipment are designated: Low voltage (600V) 500 MCM, Low Voltage (600V) 3/C 19/#22, Low Voltage Control and Power (600V) 2/C #16, 3/C #16, and Instrumentation (600V) 2/C #16, 4/C #16 shielded.

These types consist of a mixture of one, two-and three-conductor cable interfaces and appropriate shields.

Individual conductors are carried through the penetration and end in either a 60-inch or 72-inch pigtail.

2/C #16 and 3/C #16 pigtails are grouped and attached to electrical connectors (Crouse-Hinds =odel number RPC-317-160-SOIN/S08N) to provide the appropriate cable match. The connectors are located in cable trays and lie in the horizontal plane.

The cable tray runs are located essentially on the outside diameter of the polar crane shield wall to route cable to the respective instrumentation or control equipment.

The electrical penetration material which is located within containment and exposed to accident environment conditions consists of stainless steel (container) ceramic plate (con-t ductor spacer) PVC and Kerite formula (conductor insulation) and aluminum (electrical connectors).

By specification each penetration type was designed to perform under the LOCA environmental conditions of pressure and j

temperature depicted within the H. B. Robinson FSAR (shown as l

Figure 3.1.1 and 3.1.2 in this report). Test information is i

recorded in References 3 and 4.

l The CP&L Brunswick Nuclear Power Plant uses Westinghouse l

designed and fabricated electrical penetrations which are l

sinilar to those in use at H. B. Robinson.

Both are cartridge type with stainless steel sleeves and both have potting conpound seals for the internal connections of the feed-through solid copper conductors.

Brunswick penetrations utilize heat-shrink tubing for small conductors internal insulation spliced to Okonite jacketed cables forming pigtails for field cable g2 hookup.

H. B. Robinson penetrations use silicone rubber l

internal insulation spliced with heat-shrink tubing to two (2) types of jacke ted insulation cables (PVC and Kerite) forming pigtails for field cable hookup.

Both use a ceramic seal to encapsulate pigtail entry and provide an i= pervious shield l

l l

- - - ~ _. _ _ - -

L

with the cartridge sleeve. A greater degree of testing was performed on Brunswick type penetrations with results found in Reference 43.

Briefly su=marized:

Thermal cycling - 20 C to 135 C (5 cycles)

Pre-aging-524 hrs.j70C(40 years)

Radiation - 2.13 x 10 RAD Steam Test - Temperature, Pressure, Humidity and Spray (per report)

Due to the dual nature of the electrical penetrations, one side in containment the other outside, mock-up of only the in-containment area was required for testing purposes.

The test data recorded and referenced above should validate qualification of the cartridge portion of the H. B. Robinson electrical penetrations.

The electrical connectors (Crouse-Hinds Model Number ((RPC.

317-160-S0lN/S08N)) used with the penetrations consist of an extruded aluminum shell with a hard anodized finish.

The connector pins / sockets are silver-placed copper. The insert material is mineral filled diallyl phthalate with a thin vafer lM of silicone rubber provided for cealing purposes.

Mineralfilleddialgylphthgatecanwithstandradiation lM exposurebatgn10 and 10 RADS with little or no permanent degradation The silicone rubber seal wafer is positioned between two plugs of diallyl phthalate and will not be significantly affected by irradiation.

The connector proper will not be affected by normal plant life operation of forty (40) years or the added accident radiation dosage as presented in Table 1.3.3.

The aluminum shell is comprised of 6061 alloy which contains

(%):

.25 copper,.6 silicon, 1.0 magnesium and.25 chromium.

A Martin hard-coat anodized finish is applied to a depth of 1.7 - 2.0 thousandths. The alloy used experiences a weight loss of 932 =g/dm' for the first day and an average of 370 mg/dm' per day thereaf ter when completely immersed boric acid solution (pH-9) heated to 200 F.g a NaOH adjusted As the shell is anodized its corrosion resistance is improved. Additionally, the connectors will not be completely immersed in boric acid solution under spray conditions, nor will the high temperature l

be maintained for a thirty- (30) day period. Therefore, the worst case of loss of mass (..S ounce per square decimeter after 30 days) will not be realized.

Sufficient shell material vill remain to preserve connector integrity.

(2)See Appendix C to this report for reference information.

(3)WCAP 7153 Investigation of Chemical Additives for Reactor Containment Sprays.

(Reference Table 8 and Figure 9.)

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v., -,,,.

As reported by Crouse-Hinds, the anodized finish provices protection sufficient to enable specifying connector to be corrosion resistant to salt spray for 300 days (in casts per MIL C-5015D and MIL-E-4970A). Per manufacturer's installation instructions, connector will provide watertight seal and will exclude water by hese spray or stream. Luring refueling (August-October, 1980), all connectors in containment were checked and tightened to provide watertight fit.

R1 A periodic check of connector clamp and shell cover screw tightness will be established and performed to assure connectors will function for the LOCA prescribed operation time of the penetrations (see Table 1.3.3).

As the clamp seal was able to maintain connector operability after a three-hundred (300) day.

salt spray test per stated MIL SPECS, it is concluded that properly maintained clamp seals will provide chemical spray protection for the required operational times of electrical penetration connector circuits (thirty (30) minutes to one (1) day).

No significant degradation due to thermal aging will be experienced by the connector during operation plant life due tomaterialsusedindesignandforfabrication. The con:tetor design temperature range is -80 7 to 275 F and is sufficient to meet the operating and LOCA temperature range established for H. 3. Robinson.

The electrical penetrations utilize a combination of five- (5) and six- (6) foot lengths of single or multiconductor cable to connect the penetration feed-through conductors to the field lA8 cable inside and outside containment. These " pigtail" cable were installed by the manufacturer and sleeved at the penetration end with heat-shrink tubing. For selective conductors, connectors were installed while the majority of pigtail cables required butt-style splicing for field cable connection.

The cabling used for pigtails was provided by CP&L/Ebasco specification / purchase and shipped to Crouse-Hinds Company for fabrication use.

For the Low Voltage Power, (600V) electrical penetrations, 500 MCM Karite cable with HI TEMP conductor insulation was provided (see Section 3.2.4 for qualification evaluation).

For Low Voltage Control and Power (600V) electrical penetrations, 3/C #16 and 2/C #16 Kerite cable with FR conductor insulation was provided (see Section 3.2.4 for qualification evaluation). For Instrumentation (600V) electrical penetrations 2/C #16 (shielded) an'd 4/C #16 (shielded), Continental Wire Al and Cable Company cable with PVC conductor insulation was provided. No qualification data is available for this cable.

CP&L has initiated a qualification test program to determine the ability of this cable to meet IEEE 323-1974 requirements using FSAR established accident parameters.

Spare pigtails will be used and cable splices per Section 3.2.5 will be 1

utilized to maintain plant configuration during tests. Wyle Laboratories will perform the tests per Qualification Plan 543/4464/ES dated July 10, 1980. Testing and reporting will i

require thirty-five (35) weeks--af ter Receipt of Order. Major Al

}

time factor will be thermal aging to achieve forty- (40) years' operating life before LOCA testing can be performed.

After review of results, a report will be sent to NRC detailing any action by CP&L dictated by these tests.

i These PVC insulated pigtails are used for instrumentation or within circuits which must perform their functions after short elapsed time periods; therefore, their long-term operability problems should not affect plant response to accident conditions.

Results of the qualification test program will determine the l Al ultimate disposition of these pigtails.

If replacement is required, a plan and schedule for accomplishment will be included in the report already stated above.

lg; 3.2.2 Electronic Transmitters H. B. Robinsor 's original design and specification called for installation and use of Fisher and Porter electronic trans-mitter for the measurement of Pressure, Level and Flow para-meters. As stated within CPLL response to NRC IE Bulletin 79-01 and the 45-day response to NRC IE Bulletin 79-01B CP&L preference, to obtain better operation and maintenance per-formance, is to change out the existing transmitters within containment--to be replaced by Rosemounts' Model No. 1153A.

Environmental tests performed on Fisher & Porter's trans-mitters (Model No.10B2496) indicate failure occurs during the high temperature, steam / chemical spray testing stage while attempting to qualify to IEEE 323-1971 parameters.

(Reference WCAP 9157 Environmental Qualification of Safetv-Related Class y Process Instrumentation).

Qualification testing of Rosemount Model 1153, Series A, per Rosemount Report No. 3788 states that the transmitter is qualified per the requirements of IEEE 323-1971. Missing from this report is the aging parameter not required for IEEE 323-i 1971 but necessary for complete LOCA qualification.

Recent Rosamount testing to qualify a transmitter to meet IEEE 323-1974 requirements has resulted in failure. A combination of thermal aging, irradiation and chemical spray test speci-fication parameters has resulted in failed components.

The initial failed element was an 0-ring comprised of sulphur cured polyethylene rubber. This allowed steam / chemical spray i

to affect electronic components.

The 0-ring mode of failure is attributed to high temperature vs. time necessary for the Arrhenius curve time compression to satisfy aging test re-quirements.

l l

3 i

This testing failure does not preclude the use of the Rese-mount 1153A within H. B. Robinson containment as it has successfully performed within the H. B. Robinson accident parameters of temperature, pressure and radiation levels.

Transmitters located in containment will be required to perform within a maximum time period of twenty-four (24) hours following accident. 0-ring failure due to high temperatt.re should not occur during this time period. Reviewing Table C-1 l

of Appendix C, NRC IE Bulletin 79-OlB, Thermal and Radiation Aging Degradation of Selected Meterials, shows that poly ~

ethylene rubber has a potential for significant aging at ten (10) years and an allowable radiation susceptibility of 10' RADS before serious degradation occurs.

Evaluating the above establishes the need to parform periodic changeout of trans-mitter 0-rings.

j Additionally, the time span to which Rosemount will qualify its IEEE 373-1974 transmitters is ten (10) years.

To assure that listed transmitters within H. B. Robinson containment remain qqg}ified a ten- (10) year replacement cycle will be adopted i

For long-term accident mitigation, Fisher & Porter trans-mitters, Model Nos. 10B2496 and 50EP1041, located within the Reactor Auxiliary Building are used. Transmitter identification numbers are FT-940, FT-943 PT-934, PT-940 and PT-943.

As these transmitters are not exposed to the LOCA accident environment, but will see the elavated radiation levels associated with reactor coolant recirculation, qualification is limited to their radiation withstand capability.

As previously stated, Fisher & Forter 10B2496 transmitters had failed environmental testing per IEEE 323-1971 requirements and reported in WCAP 9157.

Failure occurred withing six. (6) minutes of operation when in the high temperature /high pressure / spray testing environment (Table A-7, WCAP 9157).

It is noted, i

though, and stated, that the " trip" function time of operation for the transmitters was accomplished. This portion of the test program is not relevant to H. B. Robinson use of the

$^,

listed Fisher & Porter transmitters as they are not within containment and, therefore, not required to function under the harsh environmental conditions which caused test failure.

Within the same report, it is stated that Fisher & Porter transmitters had successfully operated during and af ter irradiation testing (Tablp A-6, WCAP 9157). As only a total radiation level of 4x10* RADS were achieved, additional qualification was required to meet the radiation requirements established in Table 1.3.3.

(1) Additional design changes / improvements by Rosemount would be followed to adopt improved components or materials to minimize changeout cycles.

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- Westinghouse WCAP 7744, Environmental Testing cd[ Engineered Safety Features Related Equipment states that transmitters, identifiedbyWestinghouseasFisher& Porter,Megel10B2496, had been successfully tested to a level of 2.0x10 RADS.

As the ligted Fisher & Porter transmitters are exposed to a 1.1x10 RAD level, they are considered qualified for the application and functions stated within this report. To further identify the transmitters in use at H. B. Robinson with those tested, Westinghouse has stated that instruments used were ordered as NS (nonstandard) from Fisher & Porter.

Check of purchase order and manufacturer's fabrication instructions show that the listed H. B Robinson Fisher & Porter transmitters were supplied as NS (nonstandard).

Westinghouse has been requested to supply the specific data and/or reports associated with the testing program, and it will be available for review after receipt.

3.2.3 Motor-Operated Valves Within containment at H. B. Robinson four (4) motor operators are used for valve actuation for the listed equipment in this report. They are: V-744A and V,744B. Auxiliary Cooling System and V-866A and V-866B, Safety Inj ection System.

They are Limitorque Models SMB-00 (V 866A,B) and SMB-3, with motor brake (V-744A,B). Torque motors for V.744A&B have been wound with Class H insulation. V-866A&B Torque motors and V.744A&B l NI motor brakes are wound with Class B insulation. Model SMB-00 has a Peerless built torque motor and Model SMB-3 has a Reliance built torque motor.

Qualification testing of Limitorque motor operators was performed by Franklin Institute Research Laboratories and the test reports included in Westinghouse WCAP 7410-L, Environmental Testing of Engineered Safety Fectures Related Equipment, t

Limitorque Model SMB-Os, with and without motor brake, and Class B and Class H insulation were used during the tests.

The results are applicable to the Models SMB-00 and SMB-3 used at H. B. Robinson as differences are dimensional and in torque racing only.

l The qualification testing performed by FIRL encompasses the temperature, pressure, relative humidity and chemical spray parameters for H. B. Robinson; therefore, the Limitorque motor operations within containment are considerad qualified per these parameters for.H. B. Robinson operation.

Of concern was motor brake operation due to the results of FIRL Final Report F-C2485-01, Tests of a Limitorque Valve Operator and Motor Brake Assembly, Both with Class B, Insulation Under Simulated Reactor Containment Post Accident Steam and Chemical Environments. Failure of the motor brake with sub-

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i sequent valve operator failure was reported as occurring after seven (7) days within the test program. Performance prior to this time was recorded as satisfactory.

Since the H. B.

Robinson application of the motor brake, valve operator com-bination occurs within five (5) minutes af ter LOCA initiation, it is concluded that tae intended function of this equipment will be met by the installed equipment; no further qualifying or changeout is planned.

Radiation exposure and aging rests are described within Westinghouse WCAP 7744, Enviroamental Testing oj[ Engineered Safety, Features Related Equipment. Total irradiation to 2 x 10" RADS and a thermal aging equivalent to forty (40) years is reported.

Support data for these tests are on request from from Westinghouse and will be made available for review when received.

Outside of containment for long-term accident mitigation are additional Limitorque motor valve operators which will be exposed to elevated radiation levels only.

These are CVC.381, R2 V-860A, V-860B, V-861A, V-861B, v-863A, V-863B, V-869.

The Limitorque models used are SMB-00 and SMB-1.

No motor brakes are associated with these operators. As stated previously, Westinghouse WCAg 7744 reports a test which achieved irradiation levels of 2 x 10 RADS with no failures encountered.

CP&L has requested copies of the test data from Westinghouse, and it will be made available upon receipt. No other accident environment parameters are experienced at this location; therefore, CP&L considers these motor valve operations qualified for their intended use and location.

3.2.4 Electrical Cable The electrical equipment in containment and reported within i

che equipment list of this report is connected by either 1

single conductors or multiconductor cables.

These cables run via cable trays and conduit from the electrical penetrations l

to the equipment. Connections to the electrical penetrations are made by individual or grouped cable splices, or by elec-trical connectors. At the equipment end, formal component terminals with overall tape or crimped terminals with overall cape are used for connection.

The connectors used (Crouse-Rinds Model No. RPC-117-150-POIN/P08N) were supplied with the electrical penetrations and mounted on the matching cable during construction.

For details concerning qualification of this connector, see Section 3.2.1.

For details concerning cable splices and terminals see Section 3.2.5.

l l

The electrical cable used for equipment hookup is divided into three (3). classifications:

o multiconductor - 2/C #16, 3/C #16, 3/C 19/#22 o multiconductor - 2/C #16, 4/C #16 (single drain wire utilized as shield) o single conductor - 500 MCM The unshielded multiconductor cable is used to power the identified motor-operated valves (3C 19/#22), control the identified solenoid valves and provide limit switch outputs (2/C #16, 3/C #16). The shielded multiconductor cable is used for analog signals obtained from the listed transmitter and the listed RTD temperature elements (2/C #16, 4/C #16 shielded).

The single conductor cable (500 MCM) provides power for the containment fans (RVH-1 through HVH-4). I?g shielded cables used for containment instrumentation utilize the provided electrical connectors at the penetration end.

For instrumentation within containment, a silicon rubber conductor insulation with glass binder, an uncinned bare copper drain wire and an overall silicon rubber jacket cable is used.

The manufacturer, Continental Wire and Cable Company, used their formulated Lasulation type CC-2115.

This formu-lation has been tested by the Franklin Institute Research Laboratories under Continental Wire and Cable Company instruc-tion. Final Report F-C2935 dated, October 1970 with addendum dated November 1970, details the testing specifics which included a preconditioning (aging) period of six (6) hourg at 151 F, and a subsequent test achieved exposures of 1 x 10 RADS. Also included was a chemical spray for one hundred and twenty (120) hours. The combined data for this cable insu-lation material indicates there should be no problems asso-ciated with LOCA pressure, temperature, humidity, spray, or radiation. At this time aging is the only unknown variable.

Basically, silicon rubber cable insulatiem is designed and recommended for high temperature applications.

CP&L has no plans to conduct separate testing to further qualify this cable.

j For limit switch and solenoid valve operation, a Kerite fire-resistant conductor insulation with overall fire-resistant jacket cable is in use within containment.

Inspection of in-containment field cable hookup to limit switches and solenoid valves performed the week of August 18, SI 1980 through August 22, 1980 determined that Kerite fire-resistant conductor insulation with overall fire-resistant jacket cable is used.

l The Kerite Company has attested to the ability of this cable supplied for H. B. Robinson to withstand the FSAR LOCA conditions of temperature, pressure and radiation.

In addition, test qualification included forty- (40) year aging, borated spray and 100% relative humidity to meet IEEE 323-1974 and IEEE 383-1974 requirements. Referenced reports are:

Af FIRL Report F-C4020-1 dated March 1975.

Kerite Proprietary Engineering Memo No.178 entitled,

" Determining Temperature Ratings of Cables and Pre-aging Requirements for LOCA Simulation Tests," dated December 27.-

1974 (superseded by EM178A dated May 1, 1979).

For motor power required for valve operation, a Kerite HI. TEMP conductor insulation with asbestos fillers, nylon binder tape, neoprene treated tape, with fire-resistant jacket reinforced with a cotton-sleeve cable is in use within containment.

For containment fan power, a Kerite HI TEMP conductor insu-lation with overall fire-resistant jacket, reinforced by cotton-sleeve cable is in use within containment.

The Kerite Company has attested to the ability of this cable supplied for H. 3. Robinson to withstand the FSAR LOCA conditions of temperature, pressure and radiation.

In addition, test qualification included forty- (40) year aging, borated spray and 100% relative humidity exposure to meet IEEE 323-1974 and gg IEEE 383-1974 requirements.

Referenced reports are:

FIRL Report F-C4020-2 dated March 1975.

Proprietary Engineering Memo No. 178 entitled, " Determining Temperature Ratings of Cables and Pre-aging Requirements for LOCA Simulation Tests" dated December 27, 1974 (superseded by EM 178A dated May 1, 1979 and EM 178B dated December 1, 1979).

To provide protection for cable termination at equipment end, when no formal termination method was provided, a silicone rubber tape was used. SCOTCH _70, high temperature silicone rubber tape, is used for safety-related terminations.

This product has undergone radiation testing by the ranufacturer,MingesotaMining&ManufacturingCompany i

(3M) up to 1.0x10 RADS at 40 C temperature with no major gy l

degradation of performance.

i A more comprehensive testing program to meet IEEE 23-l 1974, requirements has been performed by Kerite Componi utilizing SCOTCH 70 tape and Kerite Cable within LOCA testing chaiber. Karite has certified the use of SCOTCH 70 as detailed in Reference Number 50.

l

To assure tape qualification for H. Bl Robinson application, SCOTCH 70 tape will be used in conjunction with test control : ables during qualification testing of the electrical penetrations PVC pigtail cable being performed at Wyle U

Laboratories. Results will be documented and available after completion of PVC cable testing.

3.2.5 Cable Terminals and Splices As no qualification information could be obtained on the curre:lt in containment cable splices to the listed electrical equipment, it was decided to change out the splices with qualified components, prescribed tools and approved procedure.

This changeout was completed during the plant refueling outage (August - October) 1980.

Individual conductor splices will utilize AMP Radiation Resistant /150 C Preinsulated Splices (#53548-1). T&B 2-way Cable Connectors for Copper Cable, 500 MCM and T&B 2-way Cable Connectors for Copper Cable, #9 AWG, The splice / connector component will be crimped to the designated conductors using the manufacturer's specified crimping tool.

An appropriate sized RAYCHEM SHRINK TUBING will he applied over the individual conductor cable splice and heat shrunk using the manufacturer's specified torch. For the two- (2) and three- (3) conductor cables after the individual conductors.

are spliced using AMP P1DG (53548-1) splices, an overall jacket RAYCHEM SHRINK TUBING will be applied and heat shrunk.

The work described above has been detailed within H. B Robinson S.E.P. Modification and Setpoint Revision Form No. M-521 (revised) and will be the means to sign off the cor.pleted work.

Original splices specified as AMP Nuclear Preinsulated Environmentally Sealed Splices (#52979) were found to be incompatible with the conductor insulation thickness of installed cable. Therefore, another butt-splice component, AMP Radiation resistant /150 C preinsulated splice (#53548-1) was ordered and installed. AMP Qualification Test Report 110-11002 dated October 1, 1978 describeg a program that included total radiation exposure of 2.0 x 10 RADS, maximum temperature of 350 F, maximum pressare of 137 PSIA and a borated chemical spray lasting four (4)

U days. To assure qualification of the H. B. Robinson in-containment splices, cable undergoing testing at Wyle Laboratories will be connected with AMP PIDG terminals Raychem thermofit (heat shrink) tubing overall per the Installation Procedure M-521, Safety-Telated Cable Splices Inside Containment. Appropriate matrix combinations of splice / cable and individual cables and splices will assure identification of any single component failure which could occur during qualification testing.

Each

component has sufficient =anufacturer-supplied test data to assure qualification by analytical means.

The opportunity to R2 obtain actual test results is available and will be used.

RAYCHEM Ther=ofic Insulation Systems (heat-shrink tubing) used to complete the replace =ent splice have been qualified per the H. B. Rcbinson accident parameters.

Franklin Institute Research Laboratories Technical Report F-C4033-3 dated Jan-uary 1975 describes a program that ineguded 40-year aging, total irradiation exposure of 2.1 x 10 RADS, maximum tem-perature of 351 F, maximum pressure of 85 PSIA and a borated spray in excess of nine days.

The results of this documented test are acceptable to CP&L that the heat-shrink tubing to be used in changeover is fully qualified.

3.2.6 Solenoid Valves As reported in CP&L responses to NRC IE Bulletins 79-01 and 79-013 (45-day report), the listed solenoid valves in containment are to be replaced by qualified equipment.

The in-place ASCO solanoid valves have not exhibited poor performance or required excessive maintenance. When manufactured and supplied, ASCO Company was not required to maintain the QC/QA procedures and programs necessary to allow traceability and certification needed for qualification.

The replacement valves are also ASCO Company equipment Model Nos. NP831665E, NPS316E35E and 206-381-2U used singly or in combination to achieve their valving function.

These solenoid valve types were included in a qualification testing program to meet IEEE Standards 323, 344, and 382. Results of this testing are published in AUTOMATIC SWITCH COMPANY. Test Report No. AQS21678/TR, Revision A, entitled Qualification Tests of, Solenoid Valves M Environmental Exposure y Elevated Temperature, Radiation, Wear Aging, Seismic Simulation, Vibration Endurance, Accident Radiation and LOCA Simulation.

The test parameters subjected the valves to a maximum temperature of 346 F, a =aximum pressure of 125 PSIA, a relative humidity of 100%, a borated spray duging tha LOCA simulation and a total radiation of 2.0 x 10 RADS. The test results are divided into two (2) parts--first the evaluation of the elastemers and coil materials and second the valve occhanisms and housing.

The elastomers and coil materials, as reported, are qualified for a 4.4 year life (includes a 10% margin figure). The valve proper is qualified for a 40-year life.

This will require the coils and elastomers to be replaced on a i

scheduled basis to maintain the serviceability of the entire valve as well.as its qualification. The proposed schedule is replacement of stated components on a four- (4) year cycle.

Replacement will be performed during the closest outage or refueling to that ti=e period.

1

With the maintaining of the replacement component schedule, CP&L considers the ASCO solenoid valves fully qualified within H. B. Robinson parameters and need do no further testing or qualifying.

3.2.7 Level switches As reported in CP&L's responses to NRC IE Bulletin 79-01 Al containment level switches (LS-1925A, LS-1925B) located within containment sump would be replaced with qualified equipment as the in-place equipment was never qualified.

As supplied, the level switches are magnetic in operation and provided incremental one- (1) foot level data as water would rise in the sump.

This equipment could operate completely submerged.

A market search did not uncover any source of qualified equipment for replacement purposes.

However, a parallel investigative effort by CP&L to meet the requirements of NRC NUREG 0578, TMI Short-Term Lessons Learned, ACRS2 Containment Water Level Indication, has concluded that there should be an analog level signal generated for combined sump and containment water level to aid in reporting and mitigating TMI type accident conditions--

if ever experienced.

The current incremental level switches, Madison Model 5602 Switch Units with Type 316 Stainless Steel Stem, 10 ft. 6 inches long, with eight (8) 316 Stainless Steel Floats and one (1)

Dry Contact Switch at each level, wired with 22AWG conductor =

with Silicone Rubber insulation, will remain in place. The.

function of these switches will be assumed by the analog system.

The schedule for completion of installation is January 1, 1981. CP&L will take no further action on these level switches in conj unction with NRC IE Bulletin 79-013, 3.2.8 Motors Within containment at H

3. Robinson included in the equipment list for the report is one (1) motor type.

This is a Westinghouse Type 685.5-S used with the containment fans.

There are four (4) f ans mounced in containment designated HVH-1 through HVH-4.

Qualification testing on a complete motor / fan assembly and on t

i individual motor elements has been performed by Westinghouse.

Results are published within WCAP-9003, Fan Cooler Motor Unit Test, 1969; WCAP-7829, Fan Cooler Motor Unit Test, 1972.

WCAP-9003 testing included:

thermal preaging to an equivalent of seven (7) years a maximum pressure of 95 psia, a maximum 3

te=perature of 315 F, and use of borated spray for thirty-five (35) hours. WCAP-7829 testing 8 included:

total irradiation of equipment / components to 2 x 10 RADS, preaging to a 40-year life expectancy.

i 1

Evaluation of the test reports concludes that the H. B Robinson accident parameters are covered by the test envelopes and parameters performed on the similar Westinghouse motor / components subjected to qualification testing.

Therefore, the cantainment fan motors at H. 3. Robinson are considered qualified.

Outside of containment, the RER pump motors are in use during long-term mitigation of LOCA conditions.

The only accident parameters experienced by these pumps / motors is radiation.

The most susceptible elements / components of the motors are covered by the testing reported within WCAP-7829.

Since the RHR pump motors are of a similar type and motor windings are Thermalastic Epoxy insulated, it is concluded that the RHR pump motor is qualified for the service intended and the environment experienced during post LOCA.

Data supporting the Westinghouse testing reported within the stated WCAPs has been requested from Westinghouse and will be available for review upon receipt.

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4.0 Conclusions The electrical equipment listed within the H. B. Robinson emergency safeguard systems and associated plant system instrumentation (Reference Section 2.0) were evaluated by equipment groups (Reference Section 3.2) and are summarized as follows:

4.1 Electrical Penetrations Containment Sleeve Sections - qualified by individual manu-facturer's test reports and similar type qualification testing.

Additional action required - None.

Conductor Pigtails (Kerite Co.) penetrations having Kerite insulated pigtail cables are considered qualified by manu-f acturer's testing program and reports (Paragraph 3.1, Reference 49),

Additional action required - None.

Conductor Pigtails (Continental Wire and Cable Co.) penetrations having PVC, conductor and jacket, insulated pigtail cables are considered nonqualified.

gg Additional action required - Separate qualification testing program has been initiated and contracted with Wyle Laboratories, Huntsville, Alabama.

Results will determine whether any further action is required. When obtained, they will be relayed to the NRC. Analysis of operating time radiation exposure concludes that the plant can continue operation until testing is completed and reviewed (LER submitted).

Current schedule calls for the test program to be completed by May, 1981 (35-week test program).

Electrical Connectors - considered qualified by analysis of materials.

Additional action required - None.

4.2 Electronic Transmitters Replacement of in-containment transmitters identified within this report has been performed within the 1980 refueling outage ( August - October,1980). At this time, no fully qualified transmitter is available for nuclear plant in-containment operation. Rosemount 1153A transmitters, qualified to IEEE 323-1971 version, were used as replacements.

82 Additional action required - A program of periodic transmitter housing 0-ring replacement (performed during yearly instrument calibration check) will provide boron spray protection capabil.itf if an accident ever occurs.

(See Pargaraph 3.2.2.)

To assure

operational capability, a ten- (10) year transmitter replacement schedule has been adopted, to be modified when Rosemount can R1 certify, by test, longer life equipment is available.

4.3

, Motor-operated Valves The Limitorque motor operators listed are considered qualified by similar type testing as reported within qualification reports available from Westinghouse and Ltnitorque.

Additional required - None.

4.4 Electrical Cable The identified silicone rubber insulated cables and the Kerite insulated cables are considered qualified by sinilar type testing as reported within qualification reports available from the manufacturers.

Additional action required - None.

(Inspection held in contain=ent August 18, 1980 through August 22, 1980 concluded no PVC field cable in use to the identified instrumentation and switches.)

4.5 Cable Terminals and Splices Replacement of in-centainment terminals and splices identified within Plant Procedure M-521-1 has been performed during the refueling outage (August - October, 1980).

Additional action required - The splice procedure and materials are being tested during the qualification testing of the pene-tration pigtail cables at Wyle Laboratories to assure compatibility l

of materials and to assure the procedure provides proper LOCA

(

protection for splices. The overall heat-shrink tubing is qualified per IEEE 323-1974 by manufacturer's test.

4.6 Solenoid Valves l

l Replacement of in-containment solenoid valves identified within this report has been performed during the 1980 refueling l

outage (August - October,1980).

The ASCO valves specified as l

replacements are considered qualified by similar type testing performed by the manufacturer and reported within available qualification reports (Paragraph 3.1, Reference 47).

84 Additional action required - Noted in the manufacturer's l

report is a certified life of 4.4 years for the coil and elastomers within these solenoid valves. These elements will l

be replaced on a four- (4) year cycle to maintain complete operar.ional capability.

t l

4.7 Level Switches Original plans for replacement of the nonqualfied containment sump level switches with qualified equipment is no longer considered necessary.

The function of level determination is being assumed by a dual analog system provided in the TMI Shcrt-Term Lessons Learned Program (Equipment will be GEMS Level Sensor - Transmitter IH36496, XM36495 and Receiver RE36562.)

The existing system will be left in place. Check of E.I.-l procedure, Incident Involving Reactor Coolant System Depressuri-zation, does not reference use of this equipment; therefore, no changes are required in this procedure when switchover is accomplished.

Additional action required - None.

4.8 Motors The Westinghouse motors listed are considered qualified by similar type testing and component testing as reported within qualification reports and documents available from Westinghouse.

Additonal action required - None.

l Where required, the plant will assure a periodic maintenance program to inspect as wel?, as replace stated elements and components.

in addition, maintenance will be performed in a manner to assure equipnent is returned to operation in its qualified configuration and installation.

I

5.0 Report Quality Assurance Basic information for NRC IE Bulletin 79-01B was initially accummulated in response to NRC IE Bulletin 79-01, issued February 8,1979.

System Flow Diagrams were reviewed and Class IE electrical equipment was listed. A compiled "Q"

list for H. B. Robinson was also reviewed to supplement the basic list. Westinghouse Instruction Book of Control and Protection Instrumentation System - Volume I - entitled " System Description and Installation" was used to confirm each item and complete the list.

To assure the list was current, a review of plant modifications and "as-built" drawings was performed at the plant site.

Also, at the site, an inspection was undertaken to obtain nameplate data of listed electrical equipment - inside and outside of containment -for use in identifying manufacturer,

=odel number and, where practicable, serial nwnbers. To complete the in-containment Class IE system's loops, a study was performed to identify the electrical penetration, by canister and pin number, associated with control of, and power for, listed equipment. Utilized were Control Wiring Diagrams (3190628) and Cable Penetration Schedules (B-19067C) to establish a list of field cables and electrical penetrations associated with Class IE electrical equipment.

Next, a search of plant documentation was made to retrieve Al Purchase Orders, Specifications, Quality Compliance Reports, and correspondence related to the listed electrical equipment.

The retrieved documents were reviewed to determine the level of testing originally performed and qualification data available to determine qualification level against FSAR parameters and requirements.

The data collected above was formulated to respond to NRC IE Bulletin 79-01.

CP&L's submittal letter (June 12, 1979) identified components which were to be replaced because of:

operational choice, lack of qualification data, need for further review to determine qualification disposition.

Upon receipt of NRC IE Bulletin 79-01B, dated January 14, 1980, CP&L organized to use the 79-01 data as a base and add the supplemental data required--all within the formats designated within the bulletin. After attending the NRC Region II Meeting held in Atlanta January 31, 1980 to clarify response to the bulletin, CP&L assigned personnel and resources to meet the 45-day and 90-day report requirements as clarified. The 45-day report essentially reformatted the 79-01 response within the designated 79-01B forms and compared qualification data accummulated under 79-01 with the requirements of the FSAR.

Differences, if any, were noted and plans for any additional qualification testing or researching were formulated.

For the report, greater emphasis was placed on effects of High Energy Line Breake outside of containment. Westinghouse's Postulated Pipe

Failure Analysis Outside of Containment Report, November 9, 1973, was reviewed and it was determined that the LOCA environmental conditions would still provide the limiting environmental parameters. As concluded in the report and H. B. Robinson Modification and Setpoint Revision 212, MSLB shielding of vulnerable transmitters (Steam Line Pressure Transmitters PT 474, 475, 476, PT 484, 485, 486, PT 494, 495, 496) was installed and verified. Due to "open-air" turbine deck construction, any external MSLB event will not result in area elevated temperature or pressures--minimizing any environmental effects on detection, or mitigation, electrical equipment.

CP&L's 45-day response to IE Bulletin 79-01B was transmitted to the NRC on March 10, 1980.

In its conclusions, commitments on changeout of:

safety-class transmitters in containment, designated solenoid valves in containment, and safety-class penetration splices in containment were again stated. Also included was commitment for changeouts to occur during the next refueling cutage at HBR.

Two items were identified as requiring further qualification investigation. They were penetration connectors and Limitorque operator motor brakes.

During the development of the 45-day report, ground work for the 90-day report was established and initial activity to add Al the necessary material and data began.

Initiated was the contact of original vendors--as identified by purchase orders--

to collect qualification data, past or current, that related to the listed safety-class equipment.

Requested were specific reports with supportive test data and/or partial data which could be additively useful in determination of guidelines established qualification. The NSSS supplier, Westinghouse, as well as " turnkey" contractor, was requested to supply qualification data in support of its FSAR statements on specific safety-class electrical equipment.

Current Westinghouse test data (UCAP topical reports) related to installed electrical equipment were reviewed for applicability and negotiations were started to purchase recent supportive data.

To aid in qualifying, H. B. Robinson Unit 2 listed components, many of which were built and installed prior to the Standards, Regulatory Guides and Codes related to qualification; the Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors (part of NRC IE Bulletin 79-01B) were utilized to establish radiation and aging parameters. Rather than use the generalized values stated in the HBR FSAR, individual calculations were made for each item presented in the 90-day report exposed to the harsh accident environment or used in accident mitigation. Appendix B, Procedures For Evaluating Gamma Radiation Service Conditions, of the DOR Guidelines was used as source material and calculations derived are included in this report as Appendix A.

Utilization l

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of this data is described in Paragraph 1.3.2.

For aging considerations, material identification within components was made and data extracted from either DOR Guidelines' supplied charts or related charts compiled by other research laboratories.

(Reference Paragraph 1.3.3, Aging) Appendix C, Thermal and Radiation Aging Degradation of Selected Materials, of the DOR Guidelines was the basic source of information.

Additional sources were researched and utilized when more definitive data was needed to cover variations of materials.

Example - 01 allyl Phaphlate radiation resistance increases with added fillers, either glass, orlon, asbestos, etc.

Appendix C of this report expanded the data needed to cover l

the H. B. Robinson connector insert material and provide qualification.

Qualification was established in four ways.

One, vendor supplied test reports completely responsive te H. B. Robinson parameters and formatted to IEEE 323, 1974 Standard, General Guide for Qualilying Class I Electrical Equipment for Nuc!. ear Power Generating Stations. A majority of the stated rep 12 cement gg components were qualified in this manner.

Two, vendor supplied reports responsive to H. B. Robinson parameters, formatted to IEEE 323, Standard 1971 version, and analyzed to meet radiation and aging requirements. Replacement transmitters were quali-fied in this manner.

Three, identified Westinghouse WCAP documents either containing detailed test reports or backup data within Westinghouse possession. Motor operators, external to containment t ansmitters, and RTDs were qualified in this manner. Fourth, vendor test data, specification information and users' test reports formed a base for analytical comparison with H. B. Robinson parameters and use of DOR Guidelines to analyze material acceptability to provide acceptable equipment qualification.

Electrical penetrations and cable were qualified in this manner.

Though not formally requested by IE Bulletin 79-013, an additional paragraph, 3.2, Electrical Equipment Evaluation, was added to the 90-day response to aid in presenting qualification conclusions, These are genus oriented and not on a plant identification i

i number basis.

Additionally required was a flood level established in containment for the LOCA accident'.

Each contributive tankage and water source was identified from reports, drawings and specif1 cations and the volumes added together to determine the total amount of water available in containment. A previously reported 2CP seal leak at HER established the 120,000 gallon-per-foot-in-containment figure, as well as the volumetric configuration of the sump and containment floor.

(Reference Appendix B) With

the calculated flood level (3.2 ft.), each electrical equipment location (height) was compared to this figure. Location was determined by height measurement above the containment floor of accessible equipment and by estimation of height within containment against known equipment in the vicinity whose location height is identified on drawings.

Height of electrical equipment close to flood level was measured exactly to assure conclusions on equipment submergence are accurate. Where achieveabic, replacement transmitters were installed in a manner to obtain greatest height above the base floor to prevent submergence of electronic compartment sections of the instruments.

Review of operational requirements concluded there was no need to reposition the pressurizer water level alarm instrumentation to avoid submergence as it was not required for accident mitigation. Additionally, a time factor will be involved as the flood height will not be realized immediately.

To assure conformity of listed equipment to mounting drawings and procedures, an in-containment inspection was performed.

Conduit and condulets were checked for cracks, separations, and improper terminations.

Electrical tape was checked for complete coverage, neat application and correct type use.

R2 Field cable to listed electrical equipment was verified to be other than PVC jacketed. Nameplate information was collected from equipment added to the master list and verified for the already listed equipment.

Cable used as pigtails for select electrical penetrations was confirmed as being PVC jacketed.

Upon confirmation, contact was made with Wyle Laboratories to perform a LOCA qualification test on the PVC pigtail cable using calculated radiation levels presented in Table 1.3.3.

A test program is underway.

Completion date is still uncertain as material identification by the manufacturer is needed to establish energization levels to determine base time and temperature to achieve Arrhenius-Curve aging.

Lowest temperature per generic material results in a thirty-five week program.

To maximize use of the test program, it was arranged to include cable splices, per installation procedure, and termination tape in the test chamber so that direct qualification data can be obtained on H. B. Robinson application. A developed installation l

matrix will assure if any failure occurs, it can be attributed to the faulty item.

Two submittals were made to the NRC relaying information required for the 90-day response to IE Bulletin 79-013.

Initial transmittal of July 7,1980 was supplemented on August 29, 1980 to provide data not available at the first submittal and to add typographical or information correction.

During the week of August 25-29, 1980, an NRC on-site inspection was held at H. B. Robinson to verify designated equipment status and to review CP&L's 45-and 90-day responses to IE Bulletin 79-013 In-containment inspection indicated no items of noncompliance and observed techniques and reviewed procedures used for changeouts being performed.

An NRC report covering this inspection (RII:NW 50-261/80-20) dated September 30, 1980 relayed the no items of noncompliance information as well as the request for added clarification of SA selected material within CP&L's 90-day response. These clarifications required additional use of Appendix C of the D0R Guidelines and selected use of data found in NUREG 0588, Appendix D Sample Calculation and Type Methodology for Radiation Qualification Dose, to aid in determination of effset of radiation levels in containment sump water on electrical equipment close to the flood level.

The added clarification material has altered a portion of the originally subritted 90-day response material. To aid in identifying the added and altered portions, a R2 marking has been made to the affected pages. A total report document is being made to meet the November 1, 1980 requirement date for IE Bulletin 79-01B information submittal.

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APPENDIX A Calculations per Appendix B of IE Bulletin 79-013 to Determine Total Anticipated Radiation

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VOLUMETRIC CALCULATIONS FOR EQUIPMENT COMPARTMENT Step 1:

Reactor Power Level = 2300 MWghft.3 i

Containment Volume = 2.1 x 10 7

i 30-day dose = 1.4 x 10 RADS Step 2:

36" Wall (Concrete Shielding)

Dose = 1.5 x 10 RADS Step 3:

Compart:nent Volumne = 2.8 x 10 ft.

Correction Factor = 0.45 0.45(1.4 x 10 ) + 1.5 x 103 = 6.3015 x 106 6

= 6.3 x 10 RADS (30-day dose)

Step 4:

0 5

1/2 hour Correction Factor = 0.09 0.09(6.3 x 10 ) = 5.7 x 10 6

5 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Correction Factor = 0.15 0.15(6.3 x 10 ) = 9.5 x 10 RADS 6

6 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Correction Factor = 0.55 0.55(6.3 x 10 ) - 3.5 x 10 RADS l

Time (hrs.)

Dose (RADS)

Do 4e + 10% Margin (RADS) i 1/2 5.7 x 10 5 6

1 9.5 x 10 1.0 x 10 6

0 24 3.5 x 10 3.8 x 10

VOLUMETRIC CALCULATIONS FOR OPERATING FLOOR COMPARTMEhT Step 1:

ReactorPowerLevel=2300MWghft.3 Containment volume = 2.1 x 10 7

30-day dose = 1.4 x 10 RADS Step 2:

Not Applicable Step 3:

6 fg,3 Compartment Volume = 1.6 x 10 Correction Factor = 0.80 7

0.08(1.4 x 10 ) = 1.12 x 10 RADS (30-day dose)

Step 4:

6 1/2 hour Correction Factor = 0.09 0.09(1.12 x 10 ) = 1.0 x 10 RADS 0

3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Correction Factor = 0.28 0.28(1.12 x 10 ) = 3.1 x 10 RADS Time (hrs.)

Dose (RADS)

Dose + 10% Margin (RADS) 0 1/2 1.0 x 10 6 6

3 3.1 x 10 3.4 x 10 l

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APPENDIX B Calculations per Appendix II to H.B. Robinson NE 10th Semi-Annual Operating Report to Determine Submergence Depth l

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6D Calculation 1

%.e Calculating the Wetted Volume of the Containment Vessel

\\',, - Vetted volu=e of the containment vessel V,

+ Volume of centainmant su=p A

- Centain=ent floor area at 228' level outside the polar crane vall 3

A

- Containment floor area at 228' level inside the polar crane vall 2

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- Height of water above the 228' level (12t" = 1.04 ft.)

y V

=V

+ (A1+A)H ev s

2 V

Ys 'can be apprenmated as the volume of a cylinder, V minus the volumes 3

of a cylinder, V and a hemisphere (the reactor vessel), V3 plus the 2

volu=es of two rectangular prisms, Vg/V5 e

O e

=

17

h".*

- (V2+V)+V4+V5 V, = V3 3

V3=Wrh 7

= 1.w.h 4

= 3.14 (7.5)227

=

28(10)(15) 3 3

= 4771 ft

= 4200 ft 72 = 7rr h V5

l'"*h

= 3.14 (7)26

= 14(10)(12) 3 3

= 924 ft

= 1680 ft 3 = i(f?rr )

3 V

= h (314)(7)3 3

= 718 ft

~ ~ ~ ~ ~~

V, = 4771 - (924 + 718) + 4200 + 1680 3

= 9009 ft V

= 9009 ft.x 7.48 gal /ft g

=,67,390 gallons A can be approxi=ated by subtracting the area included in the polar crane 3

l vall from the cross sectional area of the containment vessel l

A[

\\

2 2

A A1 = Tr2 ~T#1

<t

.. !(Y

= 3.14(63)2 - 3.14(48)2 4G I

2

= 6034 ft 1

\\

l 18

.e Calculation 2

-J Calculating the Quantity of Vater Removed Frem the Contai==ent Vessel

~

Trucked Off-Site 27,900 gel Holdup Tank (0% to 96%)

49,000 gal h ptied into A CVCS hptied into Refueling 54,500 gal Water Storage Tank (67% to 83%)

hptied to Vaste 1,600 gal

' Holdup Tank (39.5% to 45%)

Total Vater Re=oved Frem Contai= ment 133,000 gallens

~

Calculation 3 Calculating the Quantity of Vater hatied Into the Contai==ent Vessel Frem the Refue' g Vater 86,000 gal Storage Ta=k (92% to 67%)

318,000 gal

-232,000 cal 86,000 gallens Frem the Scric Acid Blender 23,238 gal 21,986 gal pri=a.J vater

+ 1,252 cal beric acid 23,238 gallens Frem the Reactor Coolant System 19,762 gal Operating Level Cc=pensgted to 200 F=43,186 gal Drain Down Level at 200 F+

-28.421 Amount spilled ento floor

- 19,762 gal 129,000 gallons Total Quantity Spilled 121,698 gal /fc 129,000 gallons 4 1.06 ft

=

l l

~

l 1 ft 120,000 gallons

=

t l

\\

l l

20 L-

(,5 A

can be approri=ated as half of the area between the missile shield and 2

the inside'of the polar crane wall.

,.. m..,. N s 3

A2 (area at 228' level) 2 2

./.,

A

. g,... ss.

2*

(T#

~I#4) 3 s'l;)-,

,/

=i[3.14(45)2_3,34(3g)]

's' s

= 2671 ft V

=V

+ (A1 + A ) Hw 2

ev s

= 9009 ft3 + (6034 + 2671) ft2 (1.04 ft)

~

3

= 18062 ft 3

3 V, = 18062 ft x7.48 gal /ft

= 135105 gallons e

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I

APPENDLT C El Extracted Information Related To Radiation Exposure of Diallyl Phthalate l

l l

I 1

9

g a

p Q

Q f

b h) 3

1. Report No.

2. Government Ac.ms.on No.

3. Recipient's Catalog No.

NASA CR-1787

4. Tet's and Sut: title

5. Report Oate p'

RADIATICN EFFECTS DESIGN HANO30CK h l-fM r

SECTICN 3.

ELECTRICAL INSUI.ATING MATERIALS AND

5. Performing Organization Coce 4

CAPACITCRS L'

7. Autnor(s)

8. Performing Orgaruzation Report No.

C'.' L. Manks and D. J. Haman L

' ~.,

?

10. Work Urut No.

g. Performing Organization Name and Mdres

[

RADWICN EFFECTS INFCP.'tATICN CENTER Battelle Memorial Institute

11. Contract or Grant No.

NASW-1568 Columbus Laboratories Columbus. Chio 43201

13. Type of Report arid Period Cavered E

12. Soonsoring Agency Naree and Address Contractor Report National Aeronautic.i and Space Ad::linistration I'

300'50"9 # '"CY 000' 9

Washingten, D.C.

2l 5/.6

15. Suopementary Notes a

t6. Abs *raet 6

i This docur:ent contains su=:a.arized information relating to steady-state b

radiation ef fects on electrical insulating aterials and capacitors. The infor-f.

sation is presented in t 'th tabular and graphical form with taxc discussion.

The radiat. ion considered includes neutrons, gama rays, and charged particles.

~

The infor tation is useful to design en61neers responsible for choosing candidate

=aterials or devices for use in a radiation environment.

l l

l

\\

~

l I

' 1, l

I 1 l f

17 Key Worca (Sw;;estr$ by Author (s))

18. Oistributton Stat 3rrent l

!adiation Ef fects, Electrical Insulators, Unclassified-Unlimited l

Cacacitors, Radiation Camage

[

'9 Ceeur.tv Camf. (cf tn;s report)

20. Secunty Camf. lof tNs pagel

21. No. of PP;es

22. Pnce' Uncla s si fied Onciassified SS

$3.00 For sale by the fJational Technicti inforrestion Ser, ice, Sonngfi:Id. Virg:nia 22151 l

1l

' ( _._.. _ __

LH E

~

F material experiences threshold damage at'a dose of S. 6 x 105 rads (C) and 5,

25 percent damage at 4. 7 x 10 rads (C). These doses are based upon 0

losses in. ultimate elongation and impact strength. Another property of poly-amide that deteriorates from radiation exposure is stiffness in flexure, which has increased between 52 and 181 percent, depending upon the nylon type, after exposure to an electron dose of 5.8 x 10 16 e/cm2 (g = 1, o Mey) at 60 C(ll). This same exposure improved the tensile strength by 49 to 107 percent. These results agree with other radiation studies which have F,

shown increases in tensile strength of 25 percent for doses over 109 rads (C).

Information on the effects of radiation on the electrical properties of polyamide is limited to results of the electron irradiation mentioned above.

I Exposure to this radiation environment produced an increase of approxi-mately one order of magnitude in the insulation resistance and a decrease of less than an order of magnitude for the dissipation factor. A decrease I

in dielectric constant was insignificant at 1 MHz and varied between 5 and 32 percent at 1 KHz, depending on the polyamide type.

~

s Diallyl Phthalate l

Diallyl phthalate with various fillers such as glass or Orlo t has

~

shown exceptional radiation tolerance for a plastic insulating material.

Little or no permanent degradation of physical or electrical properties have been observed with radiation exposures to doses of between 108 and 10 0 1

rads (C). ~ Insignificant changes are observed in the hardness and flexibility l

of this material when irradiated to these total doses. The ultimate elonga-tion and tensile strength of Orion-filled diallyl phthalate actually increased or improved with exposure to an electron dose of 5. 8 x 10 16 e/cm2 (E = 1. 0 MeV) at 60 C.

I i -

The electrical properties of diallyl phthalate such as dielectric con-stant, dissipation factor, and insulation resistance are affected by exposura to a radiation environment such as described above. The amount of degra-I dation or change in these parameters because of this exposure is of little practical significance. Permanent changes in. dielectric constant were

)l.

1ess than 6 percent while the dissipation factor recovered to below the initial value. Increases in insulation resistance during exposure are fol-l lowed by complete recovery within approximately I hour after the irradi-ation is terminated.

l -

(

l 21 l'

.l

~

.~~

_r v

9

r

~

l's I

l; U

D**D *DfTV CW o

_a E

l 1

I 1

Domage untity of Organic 4

1 Incipient to m Id Neorty cIways uscble i

Mild to modercte Often sctisfactory j

r/

ci r u-mi Moderate to severe Limited use l

3 I

ti Phenclic, gicss Icminate

!l Ptenolic, osbestos filled E

'u Phenche, unfilled Epoxy, crornatic type curing egent Polyurethane Polyester, gloss fitted

]

I il Polyester, mineral filled gi Dially! Phtholete, m;neralfilled 3j Polyester, unfilled Mylor tj I

t Silicone, gioss filled Silicone, mineret filled d

^

Silicone, unfilled l,4

+

Meiomine-formeldehyde Ureo-formcidehyde

-^

l 1

Anahne-formcidehyd e

~"

Polystyrene

^'

Acrylonitrde/butodtene/ styrene (t,SS) l Polytmide 2

T Poyvinyl chicride Poyethylene

,,,Q

{

I PJyvinyl formcl g

i Polyvinyhdene chloride 7

Polycor conote Ket-F Polytrifluorochtoroethylene a

Polyvinyl turyret wa i

Cetfulose ccetate Polymethyl methocrytote

-+ Polyomide

(

4 Vinyl chloride ccercte I

l

, Teflon (TFE) i,,

i Tetion (FEP.)

r Naturcl rutter j

4 Styrene-cutodiene (S8R)

Neocrene rutter

" 23 I

Sihcone rutter Polypropytene h

Polyvinyhdene fluoride (Kyncr 4CC)

$l e

g 4

7 s

cio y

l 10 l03 Ca 10 iCs

,o E

Gamme Oose, reds (C) 10'3 to

lo'S 10'8 10'7 10

10

Neutron Fluence, rucm (E>0.1 Mev)(*I f

2 e

2 (c) Approumate fluence (I red (C) = 4 xic nfem )

p FIGURE 3 RELATIVE RADIATION RESISTANCE OF ORGANIC 4p INSULATING.%\\TERIALS B ASED UPON CHANGES h

t l

IN PHYSICAL PROPERTIES

,[

9 fir w m. y #

=

.wp-y _ _,

L

ML19340D706

Text

4.0 CONCLUSION

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