ML20086T295

Rev 1 to EQ Documentation Package Eqdp 36.0, Boston Insulated Wire Cable

ML20086T295
Person / Time
Site:	Robinson
Issue date:	06/14/1991
From:	Bhavin Patel CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20086T214	List: ML20086T256 ML20086T286 ML20086T295
References
FOIA-95-7 36.0, NUDOCS 9508020304
Download: ML20086T295 (30)

Text

g- -

3 .

6. -

d' ENVIRONMENTAL OUALIFICATION DOCUMENTATION PACKAGE COVER SHEET H.B. ROBINSON NUCLEAR POWER PLANT '

i EQDP NO. 36.0 , REVISION 1 e

TITLE: BOSTON INSULATED WIRE (BIW1 CABLE S

EQUIPMENT TYPE: 4/C #16 AWG TWISTED SHIELDED CABLE ,

5 MANUFACTURER: BOSTON INSULATED WIRE MODEL NO.: 15948-H-004 i

NAME/ SIGNATURE

' PREPARED BY: Bhavesh Patel -% N DATE: b'N l REVIEWED BY: Peter Yandow DATE: M Y'I/ i V  :

APPROVED BY: Joel Lewis Cb e O % N- DATE: 6 I i'il  !

O I DESCRIPTION OF REVISION:

Revised entire EQDP to move test reports from Section 7 to the EQ '

Reference Files, incorporate reference file numbers in Section 1 and address revised beta dose.

i l

l 1

Rev. 1 EQDP 36.0, Page 1 of 30 9508020304 950200 Q'

\

, (tQ, PDR FOIA SUPIK95-7 PDR

[

~" ~

4

.6 e d

e List of Effective Paces Effective Paces Revision Cover Sheet 1 LEP 1 3 - 30 1 6

O Rev. 1 EQDP 36.0, Page 2 of 30

l l

a:

l

\

Table of Contents i

. i Section Title Pace Number 1 References 4 -

2 Equipment List 5 3 Qualification Analysis 7 h

3.1 Qualification Summary 7 3.2 Qualification Requirements 7  ;

3.3 Qualification Documentation Assessment 10  :

3.4 Qualification Justification Analysis 15  ;

4 Aging Data Base Matrix 25

. 5 NRC IE Bulletins / Notices / Circulars 26 ,

6 Maintenance Required to Maintain 27 Environmental Qualification  !

7 Qualification Test Reports 28 8 General Information 29 i h

v h

t l

-l,

't I

j l

i Rev. 1 EQDP 36.0, Page 3 of 30 l m

4 .y .

4 4

1

. . e Section 1 1.0 References 1.1 "Bostrad 7E Cables, Flame and Radiation Resistant Cables for Nuclear Power Plants," BIW Report No. B915, November l 1980 (EQ Ref. File No. 162). I 1.2 " Final Report on Zone Map Study for Carolina Power &

Light Company's H.B. Robinson-Unit 2 Electric Generating Plant," Patel Report PEI-TR-831006-1, Rev. B,1-28-88 (EQ Ref. File No. 73).

1.3 SAND 79-0924C, "A Study of Strong Synergisms in Polymer ,

Degradation," Clough, Gillen, & Salazar, Sandia National Laboratories.

1.4 SAND 80-2149C, " Radiation - Thermal Degradation of PE and PVC: Mechanism of Synergisms and Dose Rate Effects,"

R.L. Clough & K.T. Gillen, Sandia National Laboratories.

1.5 SAND 80-1796C, " Occurrence and Implication of Radiation Dose Rate Effects for Material Aging Studies," K.T.

Gillen and R.L. Clough, Sandia National Laboratories.

1.6 NUREG/CR-4548 (SAND 86-0494, March 86) , " Correlation of Electrical Reactor Cable Failure with Material Degradation."

1.7 " Beta Dose Inside Containment After A LOCA," EGS

, Calculation No. EGS-TR-091200-17 (2-18-91) (EQ Ref. File No. 165, Annex F).

1.8 "NRC IE Bulletin 79-01B (90-Day Report)," CP&L, Rev. 3, l

2-1-81 (EQ Ref. File No. 59).

1.9 DOR Guidelines (Enclosure 4 to IEN 79-01B).

1.10 Engineering Evaluation Eng 88-132.

1.11 Tech. Manual 728-654-13, "RVLIS Insutruction Manual,"

Rev. O.

1.12 " Inadequate Core Cooling Monitor - 86," Tech. Manual 731-449-17, Rev. O.

1.13 " Reactor Vessel Level Instrumentation System (RVLIS)

Uncertainty Analysis," CP&L 131, 8-22-84.

1.14 "RPS Instruction Manual for Cor. trol and Protection Instrumentation," Tech. Manual 728-589-13, Rev. O.

Rev. 1 EQDP 36.0, Page 4 of 30 l

$ti l l

L a

Section 2 2.0 Eauiement List The following list contains information about all cables that are qualified by this EQDP. The Sub column indicates with an "F" that the particular cable is submerged over some portion of its length during events that cause Containment flooding.

i i

1 Rev. 1 EQDP 36.0, Page 5 of 30 l Em i

n -, -

SECTION 2 EQDP 36.0 cable List Cable Destination Destination Cable Purchase IR Nuter From To Type table Model order calculation Locat sub

" TE-413 2 C2468C P8X 74 F01/11 4/C16s 15948 N-004 MBR02886 CALC 89-015 CON F

" TE 511AA C21281A JBx-1 F01/11 4/c165 15948 N-004 MBR02886 CALC 89-015 CON F

" TE-511AB C212818 JBx 5 F01/11 4/C165 15948 N 004 HOR 02886 CALC 89 015 CON F

" TE 511Ac C21281C J8x 5 F01/11 4/c165 15948 N 004 H8R02886 CALC 89 015 CON F

," TE 511AF C21281F J8x 5 F01/11 4/C165 15948-N-004 HsR02886 CALC 89 015 CON F

" ' TE 511AG C21281G Jax 5 F01/11 4/c16s 15948 N-004 HsR02886 CALC 89 015 CON F

" TE 511BA C21284A JBM-6 C10/11 4/C165 15948-N-004 Hea02886 CALC 89 015 CON TE-5118e C212848 JBM 7 c10/11 4/c165 15948 N 004 NBR02886 CALC 89 015 CON

• • TE 511BF C21284F Jax 9 c10/11 4/c165 15948 N 004 HsR02886 CALC 89 015 CON Rev. 1 EQDP 36.0, Page 6 of 30 l m.

O.

v- ,- +. se - .+...e m... .m,,,

p __._

Section 3 3.0 Oualification Analysis 3.1 Oualification Summary 3.1.1 Discussion Based on a comparison of the qualification test documentation to the normal and accident service conditions postulated for inside containment, the subject BIW cables have been demonstrated to have a 40 year qualified life for use in the RVLIS RTD circuits at HBR-

2. There are no age sensitive materials which require periodic replacement and the cable will perform its safety function during and subsequent to a DBE at the end of its qualified life. The cables are qualified to the requirements of NUREG-0588, Category I and 10CFR50.49.

3.2 Oualification Reauirements 3.2.1 Normal Service Conditions Reference Temperature: 120*F for 100% of life Ref. 1.2 Radiation: 1.1 x 10' rads gamma Ref. 1.2 (air equivalent 40 years)

Humidity: 20-90% (35% Average) Ref. 1.2 Cycling: N/A cycles per year See Section 3.4.5 Voltage: 5 VDC Ref. 1.12 Current: 2 mA Ref. 1.12 3.2.2 Desian Basis Events Source: MSLB X worst case HELB LOCA X worst case Operating Time: 30 days Ref. 1.8 Radiation: 1.5 x 10[ rads gamma See Section 3.4.7 8.0 x 10 rads beta See Section 3.4.7 Rev. 1 EQDP 36.0, Page 7 of 30 l Mi\

c.. .

i.-

e 1

S

-__ection 3 (Cont.) 4 Reference 265'F' Ref. 1.2 Peak. Temp.:

42 psig Ref. 1.2

' Peak Pressure:

Ref. 1.2 Relative Humidity: 100% maximum Ref. 1.2

$ 100% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Yes 1 N o __., Ref. 1.10 Submergence:

Yes _2L, No FSAR Section Chemical Spray: 6.1.1.2  ;

i l

'3.2.3 Functional Reauirements (Desian Basis Events) -]

Tha subject equipment must remain functional during and after a I d2 sign. basis event.

Mode: Active ,

Passive X N/A N/A Accuracy: f I

N/A N/A Response Time:  !

l See Section 3.4.5 See Section 3.4.5 Insulation l Resistance: I c

Circuit Continuity: N/A N/A 4 N/A Repeatability: N/A i Other (specify) N/A N/A l

1 -;

EQDP 36.0, Page 8 of 30 1 Rev. 1

. . . ~ _ ._

HBR-2 ENVIRONMENTAL PROFILES i

J j <00 -

j- 350 =

.00 -

• 2 64.T

• F M8R CONDITDetS ISO -

289' F f

W 200 -

152* F

,h.

. Is3 -

O u 10 0 - ,

80 I i I e 10 i so,000 1 MOUR SECL DAY SEC.

TlWE F!qure 1 HBR-2 P0 kit ACCIDENT CONTAINMENT TLi4NRATURE .-

90 =

k 40 -

E -

I 70 -

I.a 3 -

E .2 -

> 4t P$le HS A C ONDITIONS i

• 40 =

, ,0 -

to PSIG i

10 s PSIS i t i I IO e 30,000 i Hoult SEC3. DAY SEC.

TlWE Figure 2 2 R-2 POST-ACCIDENT '

CONTAINMENT PRES 5URE EQDP 36.0, Page 9'of 30 Rev. 1 Iki'5,'

I l

\

e - e 3.3 Oualification-Documentation Assessment 3.3.1 oualification Documentation 3.3.1.1 "Bostrad 7E Cables, Flame and Radiation Resistant Cable 1980.

Documentation and l

of Oualification 3.3.2 ADolicabilityof Similarity Tested / Installed Ecuiument The subject BIW cables were purchased for HBR-2 on Purchase the suppliedOrder cables No. HBR-02886.

to Test Report B915A(see letter Section from BIW certifies 8.1). are The cables addressed by this EQDP (See Section 2) 4C/#16 AWG which are Fig. 1).

similar Therefore, in material the and constructio 1.1, p. 4, the (See Reference are addressed by subject cables at HBR-2 qualification documentation contained in this EQDP.

3.3.3 _DBE Oualification Analysia 1 911 Eo. j and  !

1.

Are all peak temperature, pressure, humidity requirements enveloped during the ._1_

transient phase? l Temp. Peak 340*F 110 psig Pres. Peak R.H. 100%

Reference / Comments: Ref. 1.1, p. 15-16.

2. Are all temperature, pressure, and time requirements enveloped during the _X_

post-accident phase?

The test conditions in BIW Report No.

Reference / Comments:

B915 envelop all HBR-2 post-accident An evaluation requirements of except pressure until 30 days after the accident.

the post-accident submergence is documented in Section pressure is documented in Section 3.4.3.

l J EQDP 36.0, Page 10 of 30 Rev. 1 MM.

0

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

{ . e l

4-i XRE E9 .

tion -LT - (Cont. ) i Are margins as applied323-1974,consistent with and/or _

3. _X_ l those the DORdefined in IEEEas applicable?

Guidelines, 1

- Mcrgins:

2

'ND. of transients N/A ,

Frequency 120 times field voltage Vcitage 75'F at the peak Temperature 162% at the peak Pressure l

> 100% f Radiation

> 100% (Except post-accident Accident Duration submergence)

The test conditions in BIW Report t forNo.theB915 30 Reference / Comments: envelop containmentrequirements excep requirement andallthe HBR-2 post-accident 5 psig An evaluation _of day submergence pressure until 30 days after the accident. Section in 3.4.6. +

d the test-the post-accident submergence is documentedAn eval

- pressure is documented in Section 3.4.3. l

4. Are the functional tests as performedadequate to dem ,

requirements (includingfor the_X_

functional accuracy) equipment can be met?

See Section 3.4.5. ,

Reference / Comments:  ;

5. Was the normal and accident radiation applied prior to and/or simultaneously If yes,

__X_, - ___

with the accident simulation?- *  !

to what level?

2.0 x 10' rads gamma I See Reference-1.1, p. 15.

Reference / Comments:

does the _X_

If submergence is a requirement, test demonstrate acceptal

'6.

1 See Section 3.4.6.

Reference / Comments:

p EQDP 36.0, Page 11 of 30 R v. 1 Sih  ;

5

N

• Section 3 (Cont.) XSE E9

7. Was the requirement for chemical i spray enveloped? X l Reference / Comments: See FSAR Section 6.1.1.2 and Reference 1.1, p. 15.

l 3.3.4 Aoina Analysis Thermal

1. Does the aging analysis provided in the supporting documentation justify a 40-year qualified life at the defined service temperature? Are self heating effects adequately addressed? _)L_

Reference / Comments: Section 3.4.2 and Reference 1.1, p. 5-8. l 1

2. Was the aging analysis performed using Arrhenius techniques? X Reference / Comments: Section 3.4.2 and Reference 1.1, p. 5-8. l

3. Is the aging analysis auditable? X Referen7e/ Comments: Section 3.4.2 and Reference 1.1, p. 5-8. l

4. If any thermal aging exemptions were utilized, are they auditable? HZA HZA Reference / Comments: No thermal aging exemptions were utilized.

Radiation

5. Were gamma radiation levels met during the radiation test? X Reference / Comments: Reference 1.1, p. 15. l

6. If any radiation exemptions were utilized are they auditable? HZA HZA Reference / Comments: No radiation exemptions were utilized.

7. Are Beta radiation effects addressed? X Reference / Comments: Section 3.4.7.

Rev. 1 EQDP 36.0, Page 12 of 30 l l$,p 4

I

.- . . . as - - e 4 . w. . . . - . - -

e w

l Section 3 (Cont.' ) 121 lLQ Mechanical

8. Were the test. specimens mechanically aged? 1  ;

Cable is passive in design (i.e.,

Reference / Comments:

r- contains no moving parts) . Therefore, mechanical aging is not

[- required. ,

s 3.3.5 Test Proaram Arr.lvsis

1. Does the test sequence comply with IEEE 323-197/,? X Reference /CommerSA: .eference R 1.1, p. 2. ,

2. Are the acceptance / failure criteria clearly defined? 1 Reference / Comments: Reference 1.1, p. 10.

3. If anomalies or failures occurred during the test program, are adequate justifications provided such that the qualification is not negated? E/_A H/A Reference / Comments: No anomalies or failures occurred.

4. Are all known synergisms accounted for  ;

in the test program? X Reference / Comments: Section 3.4.4.

1

5. Are special' installation and/or maintenance requirements identified  ?

X in the test program? ,

Reference / Comments: There are no special installation or maintenance requirements in the test report (See Reference g I

1.1). -

J M

l Rev. 1 EQDP 36.0, Page 13 of 30 l - <

4 l

e

- , e I

BIW CABLE ,

TEST PROFILE Me  ;

d-

-i

==*

esp.

m

'g .

1.

w . l l

_. I w.

ee sus one, emme o es o e se as Test Results IR IR -

. Cdr $1 Cor 92 j Pressure Megohns per Megohns per l Psig 20 ft. test 20 ft. test gap. Steam length length Time F 0 7.0 x 10 8 8.0 x 10' 0 75 110 6.2 x 10 1 7.2 x 10 End of 3 bra. 340 3 0 1.1 x 10 3 1.2 x 10 4 bra. 165 0 110 .55 x 10 0 .55 x 10 7 bra. 340 0 l 05 1.1 x 10 0 1.1 x 10 10 hrs. 320 0 65 3.0 x 10 0 2.9 x 10 14 hrs. ' 300 1 1.7 x 10 1

4 days 250 20 2.0 x 10 0 1.7 x 10 1 7.6 z' 101 17 days 200 1 0 8.4 x 10 1 8.8 x 10 45 days 200 2 1 0 0.7 x 10 4.0 x 10 104 days 200 0 0 0 1.0 x 10 1.7 x 10 144 days 167 0 0.6 x 10 0 0.7 x 10 0 161 days 167 Figure 3 LOCA TEST PROFILE EQDP 36.0, Page 14 'of 30 {

Rev. 1 ets

W - . _ ,

4 e 3.4 Oualification Justification Analysis 3.4.1 General when the qualification analysis is performed Ju;tification does not clearly establish that the plant documentation This section provides the analysis necessary r quirements are met.

to cugment the qualification documentation.

Fsr BIW cable, the following areas were analyzed:

3.4.2 Qualified Post-Accident LifeDegradation CalculationEquivalency 3.4.3 3.4.4 Known Synergistic Effects 3.4.5 Functional Analysis 3.4.6 Submergence 3.4.7 Radiation Exposure Analysis 3.4.2 Oualified Life Calculation To , calculate qualified life, the non-metallic materials are assumed to follow an Arrhenius relationship. In order to calculate a qualified cnd margins, life and determine post-accident degradation equivalencythe The activation energy can be calculated from BIW cable materials. p. 8, Fig. 3, and the Arrhenius equation in th9 data in Ref. 1.1, th3 form:

' kg en 2 t3 E, =

l_. l T T 3 Where, E, = activation energy (cV) 4 eV/K) k, = Boltzmann's Constant (8.617 x 10 t = aging time at temperature T 3 tr = aging time at temperature Tr Ti = aging temperature (K) for time t3 T, = aging temperature (K) for time tr EQDP 36.0, Page 15 of 30 l l R;v. 1 l Ubb l l

on.3 (Cont.)

1.1, 350,400 hrs (40 yrs.)

Fig. 3 of Ref. ti =

= 363 K (90*C),

T1 tz

= 230 hrs T = 453 K (180'C) ring the equation gives, 350,400 4 in JO _

E, = 8. 617 x 10 _.3-_

_.l._ _.

453 363 E, = 1.15 eV ials tion tween the EPR studies and CSPEwith materthe agin

f. 1.1 does not distinguish ber cging determination and a irrgica) to be equivalent.which tical indicate the activation en(Ref. 1.6, p. 60).

test for 1440

?R/CSPE cablesitcrials are practically iden d their was aged priorI.to the LOCAThe non-metallic mate

~

life is ho cubject BIW cable Qualified

.ouro ct 155'C temperature.are listed in Table mechanisms aging:c1culated as follows:exp [ (E/ks) (1/T,-1/T ) ] 3

• Q.L. = t i wh0ra, qualified life Q.L. =

= aging time (hrs) ti

=

activation energy (eV) 4 eV/K)

E, (8.617 x 10 k,

= Boltzmann's Constant

=

aging temperature (K)

Ti a

service temperature (K) of greater than 40 T,

alifiedoflife90*C. This is very Solving the equation givestemperatureutilized a service a qu are in the RVLISlRTD due ycars assumingsince these cables lf-heating effects are negligib eThe qualified lif c nservative circuits at HBR-2 where the f these RTDs.ented in Table I.

to the low operating current oocch non-metallic material is I Page 16 of 30 i EQDP 36.0,

'N5k RCv. 1

--+

p. . .

n

.,4.3 Post-Accident

Dearadation Ecuivalency  :

exceed the equipment, the test conditions I

'or ) - the subject envelop the entire postulated t in duration. Since the non-metallic, '

'crct-case accident conditions and licnt post-accident phaseconstruction- are assumed to follow- time an and arg:nic materials relationship, of the requirements at one

• trrh nius

-Crp;rature can be transferred to another set of time-temperature

nditions using the relationship:  !

t = t, / exp [(E,/kg) (1/T,-1/T )] 3

wh ro,  ;

t = equivalent time at Ti (hours) ,

T, = reference temperature (K) t, = time at temperature T,(hours)

T, = accident temperature (K) i E, = activation energy (eV) 4 ka

= Boltzmann's Constant (8.617 x 10 eV/K) d margin by which the test conditions exceed the plant pos post-accident phase by using the following parameters: ,

T = 152*F t2 = 2400 hrs Tr

= 200'F l t3 = 1368 hrs I j

T3 = 167'F E, = 1.15 eV f

. l

-)

Th3 first four days of the test profile are considered to envelop- j th3 HBR-2 post-accident transien.t. This is very conservative since '

tho HBR-2 transient lasts less than one day. After one day, the j HBR-2 accident temperature profile remains below 153* F. By solving ths equation, the test profile exceeds the HBR-2 requirements, including required margin.  !

l EQDP 36.0, Page 17 of 30

-R v. 1 dd

- - - - . I

Section 3 (Cont.)

The test pressure profile exceeds the HBR-2 requirements for the initial four days of the post-accident period. After four days, the test pressure was reduced to O psig but the HBR-2 post-accident pressure profile remains at 5 psig until 30 days after the accident. However, the two pressure transients in the test profile with a peak pressure of 110 psig in each transient greatly exceed the HBR peak pressure requirement of 42 psig. In addition, the second pressure transient would produce more stress than a sustained pressure. Therefore, the test pressure profile can be considered to be more severe than the HBR-2 requirements.

Therefore, based on the analysis presented above, the device is qualified for the postulated accident and post-accident phases.

3.4.4 Known Syneraistic Effects Two known synergisms (Regulatory Guide 1.89, Paragraph c.5.a) have been noted for polymers such as EPR, XLPO, Chloroprene, CSPE, PE, ,

and PVC. The subject BIW cable contains EPR and CSPE. These synergisms were not addressed during the test. They are as follows:

Radiation prior to thermal aging

+

Dose rate effects A complete explanation of these known potential synergisms is provided in References 1.3, 1.4, and 1.5. The intent of the

' research documented in References 1.3, 1.4, and 1.5 was to highlight potential aging problem areas between real-time and artificially accelerated radiation aging and the sequence of radiation testing. For Class 1E equipment located in harsh environments, qualification testing to the postulated normal and accident radiation levels may not degrade the equipment to a level which is indicative of actual degradation experienced in its proposed service life.

Performing qualification testing at levels in excess of the required level is a common method to account for this synergistic effect. This device was radiation aged to a level 5.7 times (470%

tbove) the requirement. Therefore, it is judged this device was degraded to a level much greater than would be seen in service and therefore no surveillance program would be required.

l l

L 1

1 Rev. 1 EQDP 36.0, Page 18 of 30

9 _. t P

i 3.4.5 Functional Analysis .

The subject BIW cables are required to function during and after a .

DBE. These cables provide a circuit path for the RVLIS RTD circuits. These cables are rated for 600V which greatly exceeds the approx. 5 volts the RTD circuits will experience. The cables are rated for operation at 90'C which greatly exceeds the -

requirement for the RTD circuits since self-heating effects are negligible for this application.

The functional characteristic of concern for these cables is the insulation resistance. The worst IR value demonstrated during testing (Ref. 1.1) was 3.5 megohms for a 20 foot specimen. Using  :

these test values, installed cable IR is computed by the formula:

IR cable = IR spec x length spec / length cable and is equal to:

• 280 Kilohms for the TE-413-2 loop (250 ft.

installed cable length), and 233 Kilohms for the TE-511's (300 ft installed cable length assumed). ,

TE-413-2 is a 200 chm RTD (i.e., its resistance is 200 ohms at 32*F). At a temperature of 600*F, its resistance value is 440.384 +

ohms (Ref, 1.14). With 280 Kilohms shunt resistance from the cable IR, TE-413-2's effective resistance would be 439.6925 ohms. This '

would indicate a temperature of 598.28'F. Therefore, the BIW cable

, IR introduces a temperature error of -1.72*F or -0.286% at 600*F  :

true. Based upon the RVLIS uncertainty analysis performed by l Westinghouse (Ref. 1.13), a 14'F RCS temperature error at 570*F (a 2.46% error) creates a i 2.34% RVLIS level error. An additional '

-0.286% error from the BIW cable IR could introduce an additional O.27% level error, producing a level error due to RCS temperature 4

error of 2.61%. When statistically (root-mean-square method) ,

combined with the other error components, the additional error will  ;

produce less than a 0.1% level error increase above the worst case '

error of 5.1%. Therefore, the 16% accuracy targeted by  ;

Westinghouse as the RVLIS required level accuracy would ne'c be exceeded due to BIW cable IR. ,

A similar analysis for the TE-511's follows. Assuming a true reference leg temperature equal to the worst case containment temperature of 263.7'F, the TE-511's resistances are each 149.62 ohms (at 265'F, 100 ohm RTD, Ref. 1.11). With a shunt resistance i due to the BIW cable IR of 233 Kilohms, TE-522's effective ,

resistance would be 149.52 ohms. This would indicate a temperature '

of 264.5'F, a -0.5'F error. The Westinghouse analysis (Ref. 1.13)  !

shows a 10.56% level error contribution associated with a 7'F reference leg temperature error. An additional -0.5'F reference '

Rev. 1 EQDP 36.0, Page 19 of 30 M

w I

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

Tr -----_-_

e iection 3_ (Cont.)

error would increase the levell level errorerror to 10.6%.

leg temperatureStatistical combination of errors would produce a totastill well within the req of 5.105% or a 0.005% increase, RVLIS accuracy of d6%. 5 megohms for i

Based 20 ft.

upon the above analysis, the BIW cable IR of 3.is satisfactery 3.4.6 Submeroence cables supplied for use in the RVLIS However, theRTD RTD The BIW BostradB. 7E Robinson are rated for 600V. Therefore, the circuits at H. circuits operate at approximately 5 volts and 2 mA.

design rating for these cables h RTD circuits.

greatly exceeds These cables the electrical will the

. performance requirements temperaturesfor t e significantly higher ithan ted with not operate at containment containment ambient temperature since the heat rise assoc a Therefore, application is negligible. bles under this temperatures may be used for the evaluation of these ca both normal and accident conditions. temperature i for 30 The BIW Bostrad 7E cables associated with the RVLIScompen days post-accident under submerged conditions. submerged conditions to of cables to function under post-accident inside containment requires evaluation of available test dataassurance the submer determine retain sufficient if thereproperties is reasonable to perform their required catastrophic safety to cause The potential for cracking, submergence of the cable function. (creep shortout, corroF,lon) failure (ampacity, insulation submergence on required performance must be considered in addition characteristics to the effects of insulation resistance, etc.) for the cable.

Qualification testing for these cables for post-LOCA 323-1974 and the results operation are has been performed in accordance with IEEE This report did not include documented in BIW test report B915.

testing of the cables explicitly conditions. for long-term post-accident However, short term under submerged t LOCA operationsubmergence data is included in test report test B915 samples for the pos -

383-1974) for which the bend test (IEEE STD successfully withstood a 40x bend This datafollowed by 2400 in conjunction VAC with thefor 5 minutes immersed in water. can be used to provide reasonable evaluation of additional data performing their required assurance these cables are capable of function under submerged conditions.

EQDP 36.0, Page 20 of 30 Rev. 1

'Mi

%~ __

. e iction 1 (Cont.) BIW test report B915 also ption testing per n cddition to the post-LOCA bend tion test.

test,neludes However, the 52 l aging data for l CEA-S-68-516.

.g:d test cck or irradiated durationprior to the resulted at 75aC absorp in significant therma The luring the test. The first 48 weeks of this test would be 50*C (120*F).

aquiv21ent to more than 15 years of operation atis hthat52the weekIR values significant result of this absorption test fcr the cable did not vary significantly throughout tddj7.9 e to duration. With the exception of the first IR value recor ethe IR val x 10 2 Megohms at one day),

3 Megohms per 1000 feet throughout t the end 10 4.0 x It is also significant to;t.

that i

the IR value recorded at value recorded during the of the test was 3 equal to the Megohms).

These h ghesresults indicate that long-term impact on (4.0 x 10 cubm;rgence of these cables does not haveG/CR- a significantThis IR values. 1.6) which indicates the IR values for EPR cable 4548 (Ref.

insulation are predominantly temperature related.

of testing performed relates to NUREG/CR-4548 contains data and results ent building cddress how degradation of elastomeric cable ti materials cor to cetual circuit f ailure of cables inside the con a nmThe cable The same samples used f of a nuclear power plant. jacket materials.used l

in the had EPR insulation andconsideration.

CSPE (Hypalon)m2tcrials The performanceare contained i RVLIS RTD circuits under is the IR value.

• chnracteristics of concern for indicates these the IR valuesatfor cables HBR EPR Th3 data in NUREG/CR-4548 insulation are temperature dependent but recover to previous and valueswillwhen increases Therefore, it can be expected that the most decrease the as the ,

t mperature l tcmperature decreases. Therefore, the IR limiting IR values will be obtained during the should post-accident

~~_

tcmperature transient when the peak temp occurs.long-term submergence be test values which would occur during b unded by the values obtained during the post-accident conditions documented in BIW test report B915. f d in NUREG/CR-indicates that Although the samples used for the testing irradiated, documente this report re on not 4548 were also leakageexposu currents.

daterioration of IR values due to accident radiation t HBR is less ,

the order of 10 Mrad /hr leads to negligibly smallThe pof following the accident.

rad /hr immediately at HBR should 10 6 radiation exposure 4.2 x than bles.

Therefore, the post-accident rcsult in a negligible effect on IR values for these ca the potential for In addition to the effect on IR values,The primary mechanisms insulation failure must also be addressed.NUREG/CR-4548 are creep J for insulation failure identified in EQDP 36.0, Page 21 of 30 I I

Rev. 1 35  ;

l I

w - ,

e (4

cable is on 3 (Cont.) Creep failure occurs where thewith small curvature.

an edge

. cut cnd cracking. re routed in conduits

ch:d by an applied force over The RVLIS cables able.cables citu2 tion typically which occurs go where to penetration F01) a are routed 1cng vertical runs of unsupported c long vertical runs ofwith the are no potentia r c:nsideration (thoseablo trays, i h stress not conduit, points sidered and a there creep

.pported cable to Therefore, create h g creep shortout Even through is not con crc p shortout. cables.

11blo failure mode for these t failure mode for re and radiation the RVLIS hardening rtcut'is not considered 1G3, it should be noted that a credible tempera u d exposure time.

d to slow down creeping with increase it occurs in the significant failure mode when ticularly significant is a scking of liquid or condensate and concludes that is par NUREG/CR-4548 sacnca are submerged. posure ito high temperaturess on the order of r cables which in the vcro cracking will occur for short periodsking during of time ex butfromthattemperatures t me test to result 1500C)

.cra cre required for cracFor cables under no outside stress,uld not appea inga of 100*C to 140*C. The normal 50*C sculto indicate through cracks wo did not exceed 100*C. is approximately at HBR 100*C temperatureking which na tcmperature inside containment to g ccptrature This is significantly less thanwould Therefore, not result in significant crac 120*F). d of exposure.

UREG/CR-4548 conclu es

ha inside insulation containment material after would resultinsu five years

ha normal effectoperating temperaturesThe The post-LOCA peak temp in a negligible on the cable post-LOCAt i

yacro of operation. than three hours. temperature inside conta nmenThe 130*C) which lasts for less h

' tccparature profile indicates t e i t ly one day.ture excursions is not longinsulating w:uld drop below 100*C after approx ma e of the degradation duration of these post-accident tempera i ed in NUREG/CR-4548.

cntugh to result in significant m;tcrial based on test results conta nt data obtained to was be for samples wouldHowever, have this NUREG/CR-4548 thermally agedindicates only andi that f

theontes theis data present.

radiation cracking should which were cdjusted if substantial radiation agradiation ng in a rcp rt also notes thatfor thetheeffect o backgroundt in BIW report B915 negligibly small normal the be In addition, the post-LOCAofbend teswhich operation had been at 90*C then thermally aged t; rccctor.

w o conducted for samples years i  !

The successful completion of th s  !

cquivalent of more than 40 in water irradiated with 200 megarads. red in forthese5 minutes samplesdemonstrates which hadto post-ac tCot at 2400 VAC whileand irradiated immersedthSt subjected through nce that cracking ha been thermally aged,Therefore, there is reasonable assurain the failure ofl transient testing.through cracking would EQDP not36.0, result Page 22 of 30 )

R;v. 1 $21 /

~

Section 3 (Cont.)

under post-accident submergence conditions.

The corrosive effects of the containment spray and reactor coolant discharged to containment must be considered for cable insulating material subject to long-term submergence. The compounds of concern for HBR are H B03 3 and NaOH. Both the cable insulation (EPR) and jacket material (CSPE) have excellent resistance to corrosive materials according to the Handbook of Plastics and Elastomers (Charles A. Harper, 1975), and the Chemical Encineers Handbook (Robert H. Perry / Cecil H. Chilton, 1973).

Conclusion The IR value is the performance characteristic of concern for the RVLIS RTD cables at HBR-2. The test data contained in NUREG/CR-4548 indicates that the IR values for EPR insulation are predominantly temperature dependent with IR values decreasing as temperature increases. Therefore, the IR data obtained during the peak temperature transient of the post-accident test in BIW test report f B915 would envelop the conditions which would occur during post-accident submergence. Creep shortout is not a concern for the RVLIS RTD cables since these cables are routed in conduit and have no high stress points to cause creep failure of the insulation.

NUREG/CR-4548 concludes that insulation failure due to cracking would not be a concern for cables which have been in operation less than five years at temperatures below 100*C. The test results indicate that additional testing would demonstrate a much longer '

operating life for the 50*C (120*F) normal containment temperature at HBR-2. A review of material data for EPR and CSPE indicates that the containment spray and reactor coolant solutions would not have l any adverse effects on the cable insulating materials under i submergence conditions. Therefore, it can be concluded there is l reasonable assurance that these BIW Bostrad 7E cables are capable l of performing their required function for a 30 day period post-LOCA i under submergence conditions inside containment at HBR-2. l 3.4.7 Radiation Exposure Analysis The specified gamma dose (in containment) at HBR-2 is 1.5 x 10' rads The BIW cable specimens were exposed to 2.0 x(see 10, Reference rads gamma 1.7)

(see. Reference 1.1) and then successfully completed their test program, so the BIW cables are considered qualified for the worst case gamma environment at HBR-2.

The maximum unshipided beta radiation dose inside containment at HBR-2 is 8.0 x 10 rads per Ref. 1.7. The subject cables have a minimum jacket thickness of 36 mils. This outer jacket is more than sufficient to reduce the beta dose to the individual i conductors to less than 10% of the unshielded value. In addition, Rev. 1 EQDP 36.0, Page 23 of 30 l

_L .. , m

e

• g i

f

~

Section 3 (Cont.)

the BIW cable to the RTDs listed'in Section 2.0 is contained in conduit and cable trays.= Therefore, NUREG/CR-4548 allows the.

crediting of a 50% reduction in - beta radiation dose. -8.0So x. 10the maximum beta dose to.the conductors will be less jhan l rads. . The test specimens were exposed to 2.0 x 10 rads gamma at .

Isomedix (Reference 1.1, p. 10), so they are. considered qualified. '

for the beta dose they might experience at HBR-2.

a l

'l 1

.. 1 f

?

e-Rev. 1 EQDP 36.0, Page 24 of 30 4

- . , ~ ~ - . . ...

.. , e

-Section-4 4.0 Acina Data Base Matrix The subject cables are qualified for greater than 40-years service life and have no parts which require periodic maintenance or replacement.

(

't

-l i

l k?

Rev. 1 EQDP 36.0, Page 25 of 30 l

(

. ~ . ._. _ __ _ _ . _

,. . e l  ;

It

?'

t Section 5-  ;

f. 5.0 NRC IE Bulletins / Notices /Circula u  !

No IE Bulletins / Notices / Circulars were identified to be applicable '

to Boston Insulated. Wire (BIW) Cable t

a 4

1 4

4 1

Rev. 1 EQDP 36.0, Page 26 of 30 l' i

y _ ,-

. .a .

. 3 4

l t

' Section ( i 6.0 Maintenance Recuir,gj_to Maintain Environmental Oualification Device Tag: See Section 2-  ;

Manufacturer /Model/ Serial Number: BIW/Part Number 15948-H-004 I 4

t Location: Inside primary containment ,

(various).

EQ Maintenance Requirements: None Replacement Model: N/A {

Spare Components: N/A Qualification Level: NUREG 0588, Category I .i l

t 6

.. i

-a .

l i

l 1

l l

v; 1 Rev. 1 EQDP 36.0, Page 77 of 30 5 - ay b te==-6 D e opsewe oww .

~, - _ .

.~.

.e n

=., ,

' Section 7 p.

~ 7.0 Oualification Test Reoorts I

I Qualification test reports are located'in the EQ Reference Files.

p

.p.

A

'( ,

Rev. 1 EQDP 36.0, Page 28 of 30 ,

4

~t-4 -.--a+i...we .-m---- m_m . ., . , , , , __

e p; . , , : . .

5 :.

~

, 'Section 8#

g  : 8.0 General Information 8.1 . Letter from A. Sutkus (BIW) to Diia Windsor (CP&L) dated

' August 20, 1982.

is .

L. --

1.

l I-a.

l t

Y,-

Rev. 1 EQDP 36.0, Page 29 of 30

. . i

m _

e s .

-4

^~"

Section 8.1 Attachtaent 1 Page 1 of 1 BlW CABLE SYSTEMS, INC.

e. .

Parsfmitt 60510% we oltrs ee n 7M.210.* Titt a 94 0604 August 20. 1982 Mr. Duke Windsor Carolina Power & Light Company P.O. Box 1384 Hartsville. SC 29550 Gentlemen:

' CAROLINA POER & LIGHT C0., P.O. #H8R-02886 8!W JOB e20127 8!W certifies that the four items supplied to Carolina Power &

Light Company on the above mentioned purchase order meet the requirements of BIW Report #B915 furnished with the test docu-mentation. -

If you have any questions regarding this report. please contact m.

Very truly yours.

B!W CABLC SYSTEMS. INC.

M A. Sutkus Manager. Quality Control AS/md e-

%. j Rev. 1 EQDP 36.0, Page 30 cf 30

.o

Z h_ _

e /,8 " o g

UNITED STATES NUCLEAR REGULATORY COMMISSION

[ o PEGloN ll 5 l 101 MARIETTA STREET.N.W.

'f ATLANTA. GEORGI A 30323 CIRl

• • • • • January 10, 1992 Docket No. 50-261 License No. DPR-23 Carolina Power and Light Company ATTN: Mr. Lynn W. Eury Executive Vice President Power Supply P. O. Box 1551 Raleigh, NC 27602 Gentlemen:

SUBJECT:

NOTICE OF DEVIATION (NRC INSPECTION REPORT NO. 50-261/91-21)

This refers to the special Electrical Distribution System Functional Inspection (EDSF1) conducted by a team of inspectors and led by N. Merriweather from this office on September 23 - October 25, 1991. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the report.

The team assessed the design, implementation, and engineering and technical support relative to the Electrical Distribution System (EDS). The inspection consisted of a selective review of design calculations, relevant procedures, representative records, installed equipment in the plant, and interviews with engineering and technical support staff.

The inspection concluded the Robinson EDS was adequate to perform its intended safety function, EDS material condition was good, and engineering and technical support for the EDS was adequate. There is limited margin for future load growth on the emergency busses. The team noted that you were aware of the limitations of the EDS and had mechanisms in place to monitor and control load growth. It was the team's understanding that a project had been initiated to investigate possible alternatives to increase the emergency system margins.

The team considered this action to be a strength.

The team had several findings in the areas of design, maintenance and testing. While the findings do not indicate programmatic problems they should be given appropriate attention for timely resolution. These concerns are described in the enclosed report details and summarized in Appendix A. We request that you respond in writing to each of these findings (except for finding number 5) within 60 days from the date of this report. Your response should state what action you intend to take and when the action will be complete.

Based on the results of this inspection, certain of your activities appeared to deviate from commitments to the NRC, as specified in the enclosed Notice of Deviation (Notice). We are concerned about the oeviation because your sampling methodology may compromise your ability to detect diesel fuel oil degradation.

, 1p.g.J } {

Carolina Power and Light Company 2 January 10, 1992 r You are required to respond to this letter and should follow the instructions In your specified in the enclosed Notice when preparing your response.

response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of ,

this letter and its enclosures will be placed in the NRC Public Document Room.

The responses directed by this letter and the enclosed Notice are not subject i to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No. 96.511.

Should you have any questions concerning this letter, please contact us.

Sincerely.

roe. Albert F. Gibson, D ctor 1

Division of React Safety i

Enclosures:

I

, 1. Notice of Deviations

2. NRC Inspection Report cc w/encls:

C. R. Dietz, Vice President Robinson Nuclear Project Department H. B. Robinson Steam Electric Plant P. O. Box 790 Hartsville, SC 29550 R. H. Chambers, Plant General Manager H. B. Robinson Steam Electric Plant P. O. Box 790 Hartsville, SC 29550 Heyward G. Shealy, Chief Bureau of Radiological Health Dept. of Health and Environmental Control 2600 Bull Street 1 Columbia, SC 29201 l (cc w/encls cont'd - see page 3) i

" e"---

,e Carolina Power and Light Company 3 January 10, 1992 (cc w/encls cont'd)

Dayne H. Brown, Director Division of Radiation Protection N. C. Department of Environment, Health & Natural Resources P. O. Box 27687 Raleigh, NC 27611-7687 McCuen Morrell, Chairman Darlington County Board of Supervisors County Courthouse Darlington, SC 29535 Mr. H. Ray Starling Manager - Legal Department P. O. Box 1551 Raleigh, NC 27602 H. A. Cole Special Deputy Attorney General State of North Carolina P. O. Box 629 Raleigh, NC 27602

. Robert Gruber Executive Director '

Public Staff - NCUC i P. O. Box 29520 Raleigh, NC 27626-0520 J. D. Kloosterman, Director i Regulatory Compliance H. B. Robinson Steam  ;

Electric Plant  ;

P. O. Box 790 i Hartsville, SC 29550 '

I i

l i

l

_ m-F.-

a L ENCLOSURE 1 l

NOTICE OF DEVIATION  !

Carolina Power and Light Company Docket No. 50-261  ;

H. B. Robinson License No. DPR-23 j During an NRC inspection conducted on September 23 - October 25, 1991 a deviation from a written commitment was identified. In accordance with the i

" General Statement of Policy and Procedure for NRC Enforcement Action", 10 CFR  !

Part 2,' Appendix C, (1990), the deviation is listed below: 1 Licensee letter dated May 14, 1990,- QA REQUIREMENTS REGARDING DIESEL GENERATOR FUEL OIL, committed to Regulatory Guide 1.137, Fuel Oil Systems 1 for Standby Diesel Generators, position C.2, which endorsed ASTM-  :

D270-1975, Standard Method of Sampling Petroleum and Petroleum Products.

- ASTM ~ D270-1975 requires fuel oil samples to be representative of the stored volume being sampled and provides. methods for acquiring the representative sample. y t

Contrary to the above, the licensee deviated from the above commitment' in that diesel fuel oil storage tank sample methodology does not implement i

,- the methodologies provided in ASTM D270-1975. No justification i

demonstrating equivalency between the ASTM methodologies and- that d

, practiced by the licensee was documented.  !

.i Please provide to . the U.S. Nuclear Regulatory Commission, ATTN: Document- 3 Control Desk, Washington, D.C. 20555, with a' copy to the Regional  !

Administrator, Region II, and if applicable, a copy to the NRC Resident ,

Inspector, in writing within 30 days of the date of this Notice, the reason for ,!

s the deviation, the corrective steps which have been taken and the results l achieved, the corrective steps which will be taken to avoid further deviations, i and the date when your corrective action will be completed. Where good cause  ;

is shown, consideration will be given to-extending the response time. '

FOR THE NUCL REGULATORY COMMISSION x

Albert F. Gibson, Director Division'of Reactor Safety jl Dated at Atlanta, Georgia j this 10th day of January 1992

'l

,,g am N$

. l s.

[.

M[ 8 4#

0,,

o UNITED STATES NUCLEAR REGULATORY COMMISSION REGION 11 5 g 101 MARIETTA STREET, N.W.

• ! ATLANTA, GEORGI A 30323

%,...../

Report Nos.: 50-261/91-21 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.: 50-261 License No.: DPR-23 Facility Name: H. B. Robinson 2 Inspection Conducted: September 23, 1991 - October 25, 1991 Team Leader: .

N.'Merriweather, Team Leader / 72 ,

Date Signed Team Members: R. Moore, NRC J. Dick, AECL L. Garner, NRC L. Maggio, EPM M. Miller, NRC J. Lindley, AECL Approved by: W hL /2 M. Shymlock, Chief s / 'Date Signed Plant Systems Section Engineering Branch Division of Reactor Safety

SUMMARY

Scope:

This special, announced inspection was conducted in the areas of design of  !

electrical systems and related engineering activities. NRC Teicc*ary Instruction 2515/107, " Electrical Distribution System Functional Inspection (EDSFI)", issued October 9, 1990, provided guidance for the inspection Results:

The electrical distribution system at H. B. Robinson was capable of performing its intended functions under normal and postulated accident conditions.

In the areas inspected, three unresolved items and one deviation were identified. The unresolved items involved: marginal service water flow rate to diesels for 110% power (Finding 91-21-02), E1/E2 equipment room ambient conditions' not evaluated (Finding 91-21-03), and failure to establish surveillance procedures which adequately incorporate EDG loading requirements hYYTlW

4 .

2 stated in the Technical Specifications (Finding 91-21-09). The deviation (Finding 91-21-05) involved a failure to implement procedures for sampling the ,

EDG fuel oil in accordance with a previous commitment to the NRC.

Other team findings were in the areas of design calculations and documentation, electrical maintenance practices, and testing. A summary of team. findings is provided in Appendix A.

The team considered the licensee's design activities to be proactive in pursuing a design concern related to analyzing the fast bus transfer scheme and this was considered a strength.

l l

l 1

I I

l

. i l

l I

l

W .--

i.l  !

i n

N. I i

t h  :

i i

TABLE OF CONTENTS  ;

.l PAGE-EXECUTIVE

SUMMARY

................................................... 1

1.0 INTRODUCTION

.................................................. 1 .

2.0 ELECTRICAL SYSTEMS ............................................ 2 2.0.1 EDS Description ................................... 2 2.0.2 EDS Review Scope .................................. 3 i 2.1 Conclusions ............................................. 4 2.2 Offsite Power ............................................ 5  ;

2.2.1 Grid Stability and Reliability .................... 5  ;

2.2.2 Degraded Grid Voltage Protection .................. 6 2.3 Medium Voltage and Safety-Related 480 VAC Systems ........

7  !

2.3.1 General Design Criteria (GDC-17) .................. 7 j 2.3.2 Validation and Verification of Computer Programs .. 7

' 2.3.3 Fast Bus Transfer ................................. 8 2.3.4 Cable'Ampacity Ratings ............................ 8 2.3.5 Short Circuit Calulations ......................... 8 t

2.3.6 Voltage Drop Calculations ......................... 9' 2.3.7 Load Capacity Calculations ........................ 10  :

2.3.8 Control Circuit Voltage Drop ...................... 10 ,

2.3.9 Busses El and E2 Circuit Breaker Coordination ..... 10  ;

2.3.10 Medium Voltage Protective Relaying ................ 10 g- 2.4 Emergency Diesel Generators .............................. 11 2.4.1 Static and Dynamic Loading Analysis ............... 11 2.4.2 EDG Protection ....................................

11 2.4.3 EDG Load Sequencing System ........................ 12 2.5 Class IE Low Voltage AC Systems .......................... 12 2.6 Class IE 125 VDC System .................................. 13 2.7 Containment Electrical Penetrations ...................... 13 2.8 Grounding of AC Systems .................................. 14 3.0 MECHANICAL SYSTEMS ............................................. 14

-)

3.1 Conclusions ...............................................

3.2 EDG Loading ...............................................

14 15 3.3 EDG Support Systems ....................................... 16

-3.4 Plant Service Water Interface ............................. 16 3.5 Heating, Ventilation, and Air Conditioning (HVAC) ......... 17 3.6 EDG Fuel Oil Storage and Distribution System .............. 18 3.6.1 Inventory and Fuel Consumption . . . . . . . . . . . . . . . . . . . . 18 1

3.6.2 Miscellaneous Fuel Oil Issues ..................... 18 i l

1 4.0 MAINTENANCE, TESTING, CALIBRATION, AND CONFIGURATION CONTROL .. 19  !

4.1 Conclusions .............................................. 19  !

4.2 Eq u i pme n t Wa l kd own s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 '

4.3 Equipment Maintenance, Testing, and Calibration .......... 21

)

i l

y _ . _ . _ _ _ . -

s . o. .

4 2

-4.3.1 Preventive'and Predictive Maintenance ............. 22 4.3.2 MCCs, Protective Relays, and Swttchgear ........... 22 4.3.3 EDG Surveillance ..........;....................... 22' 5.0 ENGINEERING AND TECHNICAL SUPPORT ............................. 24 5.1 Conclusions .............................................. 24 5.2 Organization and Staff ................................... 24 5.3 - Problem Identification and kesolution . . . . . . . . . . . . . . . . . . . . 25 5.4 Routine Plant Activities ................................. 25 5.5 Modifications ............................................ 26 6.0 EXIT MEETING ....................................,,,, 27

-Appendix A: Findings Appendix B: Acron'yms and Abbreviations Appendix C: Persons Contacted l

d

)

1 EXECUTIVE SUMAst An Electrical Distribution System (EDS) functional inspection was conducted at the H. B. Robinson nuclear station over the period September 23 through October 25, 1991. Led by Region II, the team was comprised of Region II inspectors, the senior resident inspector, and supplemented by contracted design engineers.

The team reviewed the Robinson EDS design with regard to compliance with regulations and licensing commitments, design engineering standards, and accepted engineering practices. The process involved an examination of the application of EDS equipment in total system design from the non-safety swi tchyard to the safety related load; a review of the as-built material condition of the EDS; a review of EDS maintenance and testing activities; a review of the design and operation of EDS mechanical support systems; and an assessment of the engineering and technical activities related to the EDS.

' The team concluded the Robinson EDS was adequately designed and maintained to accomplish its intended safety function. EDS material condition was good, and engineering and technical support for the EDS was adequate. There is limited margin for future load growth on the emergency busses. The team noted that the licensee was aware of the limitations of the EDS and had mechanisms in place to monitor and control future load growth. It was the team's understanding that a project had been initiated to investigate possible alternatives to increase the emergency system margins. The team considered this action to be a strength.

The team concluded that the Design Basis Reconstitution Program in conjunction with the PRA was a strength in resolving issues related to GDC 17 and single failure criteria. Although the grid load flow studies were not developed with controls equivalent to the calculation control program, the team concluded the studies were a strength. The studies provided the dispatcher adequate information on capacity and configuration requirements to support the switchyard voltage at Robinson Unit 2 for all accident conditions, thereby supplementing the intent of GDC 17 requirements for offsite circuits.

The team had several findings in the areas of design, maintenance and testing.

The findings did not indicate programmatic problems and had only limited safety significance, however, they should be given appropriate attention for timely resolution. These findings are discussed in the following paragraphs and summarized in Appendix A.

The team identified that a mis coordination existed on certain non-safety breakers in the 120 VAC instrument bus system which could render safety-related buses inoperable under a seismic event. The team noted that the licensee was well aware of current industry issues and NRC Notices regarding fast bus transfer schemes and had taken action on the concern by authorizing a bus transfer dynamic simulation study. The team considered this action te be a strength. The team noted that the available service water flow to the ED4 heat exchangers was less than recommended by the vendor. This could impact EDG capacity with high lake water temperature and EDG load greater than 2500 KW.

r N~  !

ii l

The team observed that the analysis was nonconservative with respect to modeling of the EDG heat exchangers. This is considered to be an unresolved '

item.  :

The team identified the safety-related electrical equipment room ambient temperature conditions had not been adequately analyzed. This is an unresolved j item. The condition of the underground fuel oil piping was indeterminant due ,

to the failure of the galvanic corrosion protection system. Two examples of ,

i poor maintenance practices regarding sealing of spare cable ends and labeling electrical equipment were identified. The team identified problems with MOVs ,

which involved a potential undetectable MOV failure mode that may not have been previously analyzed and two undersized motor starters in the RHR system for MOVs 744A and B. The team also identified that the EDGs had not been tested at nameplate rating (3125 kva) as required by T.S. In a dissenting opinion. l licensee management stated their belief that the applicable operational t surveillance test procedures tested the EDGs as required by Technical Specifications. This is considered to be an unresolved item.

. Within the areas inspected, one deviation was identified. The deviation involved the licensee's failure to implement a previous commitment to industry standards '

relating to EDG fuel oil storage tank sampling. ,

Not withstanding the above, the team concluded that the Robinson EDS was ,

operable, E&TS was adequate to support EDS activities, and adequate design controls existed for EDS. ,

j

_a _ _

l

1.0 INTRODUCTION

Previous NRC inspections of Nuclear Power Plants and various licensee event reports have identified conditions in the electrical distribution systems at various operating plants that could compromise the design safety margins of the plants. This resulted in part from a lack of proper E&TS, which resulted in the introduction of various design deficiencies during the initial design or during subsequent design modifications of the station EDS. Examples of some of these deficiencies were unmonitored and uncontrolled load growth on safety-related buses, inadequate modifica-tions, technically inadequate calculations, incorrect facility configuration, inadequate testing, improper application of commercial grade components, lack of fuse control, and improperly installed electrical connections.

The primary objective of this inspection was to assess the capacity of the H. B. Robinson EDS to perform its intended functions during plant operating and accident conditions. A secondary objective was to assess the capability and performance of the licensee's engineering organization in providing engineering and technical support for EDS activities.

The team reviewed the Robinson EDS design with respect to regulatory requirements, licensing commitments and pertinent industry standards.

The EDS components reviewed included the .EDG and auxiliary support systems; the offsite circuits from the 115 and 230 kV switchyard, the startup and unit auxiliary transformers, the 4.16 kV switchgear and related equipment; the 480 VAC switchgear, load centers and motor control centers, the 125 VDC batteries, chargers and distribution systems; the 120 VAC distribution systems, HVAC for EDS equipment spaces; protective relaying and coordination; AC grounding and penetration protection.

Additionally, the mechanical systems which are required to support the EDG were specifically examined. These included the air start system, lube oil system, fuel oil system, and water cooling system.

There were two unique aspects of the Robinson EDS. The first was the use of 480 VAC emergency buses and 480 VAC EDG. The second aspect was that Robinson was not originally designed to meet GDC-17 with respect to 2 independent offsite circuits supplying the emergency bus. The licensee had implemented measures to meet the intent of GDC-17.

Within this report FINDINGS are identified and are defined as follows:

FINDINGS are facts or conclusions related to how well the electrical distribution system meets its intended function. FINDINGS may indicate a requirement or an accepted industry practice that was not fully implemented. FINDINGS may indicate discrepancies or omissions in documents where these problems could credibly result in the intended function being compromised. The licensee's working knowledge of the design as well as their control of design documents may be the subjects of FINDINGS. FINDINGS typically make statements about the need for corrective actions.

y

r. .  :

l 2 '

2.0. ELECTRICAL SYSTEMS 2.0.1 EDS Description  !

The Robinson Nuclear Station was connected to the CP&L high voltage grid ,

transmission system through two switchyards. Electrical energy generated by the. main generator at 22. kV is transmitted to .three single phase t transformers by an isolated phase bus duct. These transformers step the voltage up to 230 kV and are connected to a 230 kV switchyard. The 230 kV '

switchyard is connected to five outgoing lines and to an adjacent 115 kV.

switchyard through two 300 MVA auto-transformers. The 230 kV switchyard  !

is of the breaker and a half design and the 115 kV switchyard is a split '

bus design incorporating two bus sectionalizing breakers. The 115 kV. ,

switchyard has three outgoing lines. The coal fired HBR-1 and a gas j turbine generator are located adjacent to the nuclear unit but are not considered part of the off-site system. The supply to the station t auxiliaries is provided by a 22 kV/4.16 kV unit auxiliary transformer-  :

connected to the main generator isolated phase bus and by a 115 kV/4.16 kV  :

, start up transformer connected to the 115 kV switchyard.

During normal operation, the main generator through the unit auxiliary transformer ' supplies power to four of five 4.16 kV buses and the 115 kV-offsite system, through the startup transformer, supplies power to the remaining bus. Following a main - generator trip these 4.16 kV buses '

normally energized by the unit auxiliary transformer are automatically  ;

transferred to the startup transformer. 1 The two Class IE 480 VAC ' safety related buses (El & E2) are normally connected to separate 4.16 kV buses via stepdown transformers. Bus E2 is ,

connected to a 4.16 kV bus which is energized by the 115/4.16 kV startup  ;

transformer and bus El obtains power from a 4.16 kV bus energized by the '

22/4.16 kV unit auxiliary transformer. t Two onsite safety related emergency diesel generators are connected via output circuit breakers to the 480 VAC safety related buses. Each EDG has a continuous rating of 2500 KW and a short term rating of 2750 KW at 0.8 power factor. The continuous rating is based on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duty cycle while the short term rating is based on one 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> operation in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. The EDGs are capable of starting a single 900 KW motor and ,

produce power at 480 VAC. They are connected to redundant class IE buses.

If a loss of bus voltage occurs the EDGs are automatically started and their circuit breakers closed. Blackout loads are automatically recon-nected to the buses in sequence. '

1 The Class IE 125 VDC system is split into two groups; each group comprised of one battery, two battery chargers, one DC motor control center and two DC panel boards. The batteries are located within a single room with a metal protective barrier between them. The group 'A' battery has a rating of 525 A-H and the group 'B' battery a rating of 170 A-H. Based on load  ;

profiles each battery has sufficient capacity for 60 minutes of operation under emergency conditions without re-charging. j l

w 3

Two class IE 120 VAC main instrument buses are associated with each redundant train. These buses supply secondary buses which have both safety and non-safety loads connected through 1solating devices. On each train one main bus is supplied from a e asst 1E 480 VAC motor control center through a 7.5 kVA constant voltage transformer and the other main bus is supplied from a DC to AC inverter, also rated at 7.5 kVA, which is connected to a class IE 125 VDC battery. An alternate supply to each bus is provided from cne non-class IE 480 VAC MCC. Transfer to this supply is manual only and is provided for maintenance purposes. No automatic transfers to alternate sources are provided for these buses.

2.0.2 EDS Review The team reviewed the adequacy of offsite power sources that supply power to the EDS. Calculations and analyses were reviewed to insure that electrical power of acceptable voltage, current, and frequency would be available to safety related equipment supplied by the EDS. The design ratings were compared to the equipment ratings to verify that acceptable margins existed. Additionally, plant walkdowns of the offsite and onsite power sources were performed to verify that equipment nameplate ratings and the installed configuration corresponds to the design requirements and is in agreement with the facility documents.

The non safety related medium voltage 4.16 kV and 480 VAC safety buses and their connected loads were reviewed to assess voltage adequacy, load current and short circuit current capabilities, equipment ratings,

. protective relaying, circuit breaker coordination, cable sizing, and voltage relay applications. Also reviewed were load flow studies, testing and surveillance, General Design Criteria 17 compliance with single failure criterion, adequacy of the fast bus transfer scheme in terms of any effects on the safety systems, and applicable separation requirements.

The Emergency Diesel Generators were reviewed to assess the EDG sizing adequacy and the transient and steady state loading capability. The safety-related 120 VAC and 125 VDC systems were reviewed for short circuit capability, loading, minimum and maximum expected operating voltage, and equipment sizing and ratings (e.g., batteries, chargers, transformers and inverters).

The specific components and portions of the EDS reviewed and the team's conclusions are discussed in the following paragraphs.

2.1 Conclusions The team concluded that the 115/230 kV system provided stable off-site power to the plant 480 VAC safety related buses. The 480 VAC safety-related distribution system was adequately designed. No operability concerns were identified by the team. The EDGs were adequately sized, however, the licensee's EDG load studies indicated loading in excess of the continuous rating in some cases.

I 1

m ____

4 Electrical calculations associated with the Class IE 480 VAC system demonstrated that there is sufficient capacity and voltage level to support the necessary safety-related loads. Equipment was rated properly for short circuit duty. There was limited margin for future load growth on the emergency buses. The licensee was aware of load growth limitations and had established a mechanism for evaluation, monitoring, and control of future load growth. Additionally, a project was initiated to investigate possible actions to increase the emergency system margins. The team considers the licensee's action to be a strength.

Protective relays and relay coordination were satisfactory. The licensee's actions in regard to the fast bus transfer dynamic analysis indicated that the licensee is keeping abreast of industry issues involving the EDS and is seeking solutions to identified problems.

The team identified an occurrence of inadequate breaker coordination in the 120 VAC instrument bus system. A seismic event could cause two non-safety loads in both trains to short circuit which could lead to the de-energizing of both instrument buses. This was considered to have minor safety significance.

The team identified minor deficiencies in the 120 VAC calculations related to the establishment of the loads on the instrument buses.

2.2 Offsite Power

. 2.2.1 Grid stability and reliability The team reviewed the licensee design to assure a reliable source of offsite power to the emergency buses. The stability of the offsite power source to support the plant during maximum expected loads on the transmission system was adequate. In cases where the grid conditions potentially compromised the reliability of the offsite power supply to onsite emergency buses, due to lack of adequate grid voltage or capacity, appropriate contingency measures had been established to ensure reliable power to the emergency buses. These measures are discussed . n the following paragraphs.

To assure that the transmission system was maintained in a configuration to support the voltage at Robinson the Nuclear Engineering Department had a formal agreement with the Transmission and Distribution Department to conduct engineering studies of the transmission system and develop voltage schedules for various network scenarios and configurations to maintain proper of f site capacity levels. These conditions were proceduralized in System Planning and Operations Procedure DTRM-GP-22 utilized by the system dispatcher. DTRM-GP-22 provided the dispatcher guidance for eastern area loads up to 9000 MW. The eastern area peak was approximately 8500 MW set on July 23, 1991. In the event that adequate voltage and capacity to support Robinson could not be achieved the Transmission Department procedures required the dispatcher notify the plant operator. Upon notification that the voltage could not be maintained the plant offsite

l i

5 i

source would be declared inoperable and the emergency bus would be loaded on the diesel.

The operator's actions were described in Abnormal Operating Procedure A0P-027, Operating with Degraded System Voltage. In addition, CP&L established a working committee which included both NED and T&D to review operating procedures and actual operation of the system and if necessary make recommendations for changes. The team considers the licensee's formal interaction with Transmission and Distribution and the resulting procedures and schedules developed to assure a reliable offsite power i supply to be a strength. l The team reviewed the transmission grid analysis which established minimum and maximum switchyard voltages to verify adequate voltages were available for safety-related components. It was determined that the minimum switch-yard voltage required to support a LOCA had been established by NED as 112.9 Kv (see calculation RNP-E-8.002, Rev. 1). For margin,113.7 kV was used by the transmiss' ion department as the minimum allowable system voltage. They have taken this value, performed analysis, and established requirements (operation of capacitor banks and/or operation of the Darlington Plant IC turbines) to assure this voltage will be met. These requirements were formalized in system dispatcher instruction DTRM-GP-22.

The maximum expected high voltage levels occurred during light-load grid conditions. Several factors tended to raise the voltage levels seen on the power system during light-load periods. Examination of less typical combinations of factors demonstrated the Robinson 115 kV bus voltage may reach 119 kV. During light load periods (spring / fall in the early morning) high voltages have been experienced on bus E2. Bus E2 is normally connected to the grid via 4.16 kV bus 3 and the Start-up trans-former. Most other buses were powered from the generator via the Unit Auxiliary transformer.

Procedure AOP-031, Operation with High Switchyard Voltage, was established to avoid these overvoltages. If high voltage was  ;

present on bus E2, the AOP required for a realignment on the 4.16 kV system and/or increasing load on bus E2 to increase the loading on the SAT, thereby reducing voltage on bus E2. Other means of reducing these overvoltages, such as reducing the Unit minimum MVAR output limit, were being examined.

The team investigated the minimum number of transmission lines required to be energized to assure plant voltages during a LOCA. The team concluded  ;

there was an adequate number of transmission lines supporting Robinson Unit 2. The number of transmission lines required to support a LOCA was  !

I dependent on a variety of system conditions such as system load, operating plants, energized transmission lines, and energized capacitor banks.

The Darlington IC electric generating plant was the most important source l for assuring adequate voltage during a system peak particularly if Unit 1 {

and/or the capacitor banks were unavailable. Due to this apparent reliance on the Darlington plant, the team was concerned if the trans- j mission network would support Robinson Unit 2 without Darlington. In i l

1

y -.

6 response the licensee performed an analysis assessing the impact on Robinson Unit 2 from postulated failure of Darlington 230 kV bus and its ability to supply power from the Darlington plant and other transmission sources connected with this bus. Under this condition, the licensee demonstrated adequate voltage would be available in the switchyard in the event a LOCA were to occur at Unit 2. This analysis resolved the team's concerns on the issue.

2.2.2 Degraded Grid Voltage Protection The team reviewed the degraded grid voltage protection scheme and was concerned with the protection relay setpoints. The degraded grid voltage relays were set to disconnect the 480 VAC emergency buses from offsite power and transfer the buses to their associated emergency diesel generators at 415 plus or minus 4 Volts. The problem with this setpoint was it allowed operation of continuous-duty motors at voltages less than their minimum rating (typically 90 percent of nameplate rating). The DBR determined that the degraded grid voltage setting was originally based on minimum contactor pick up voltage requirements at the MCC level. As a result the licensee established operator intervention under abnormal operating procedure A0P-027, when responding to a low voltage alarm at 115 kV (plus or minus 1 kV). The operator was procedurally required to monitor the 480 V emergency bus voltage and transfer the buses to their respective diesel generators should the voltage continue to decrease to the analyzed limit for maintaining minimum rated voltage at safety-related ,

motors. l The licensee was relying on manual operator action for a function that should be an automatic relay function. This was an interim corrective action and the licensee has been communicating with NRC in development of long term corrective actions. The licensee was proceeding to alleviate tne need new for operator intervention by replacing degraded grid relays with relays.

to dropout ratio.

The new relays support a setpoint increase with lower pickup Also, the licensee was performing an analysis to determine an appropriate setpoint upon which a TS Change Request will be submitted to the NRC. The team concluded this issue was being adequately addressed.

2.3 Medium Voltage and Safety Related 480 Volt Systems 2.3.1 General Design Criteria 17 Certain provisions of GDC-17, such as two independent circuits from the transmission network to the onsite electrical distribution system and the current definition of an electrical systen. single failure, were not ,

incorporated in the design of HBR-2. The licensee's DBR Program in conjunction with the PRA was an acceptable method to identify and resolve issues related to GDC-17 and the single failure criteria. Examples where )

these methods were applied were reviewed by the team and found to' be acceptable.

The licensee's transmission system load flow studies provided

.l

7 the dispatcher with the necessary information on capacity and configura-tion requirements to support plant operations under accident conditions.

2.3.2 Validation and Verification of Computer Programs The team reviewed the results of the licensee's validation and verification analysis for the computer software used by design to perform various electrical calculations (e.g., short circuit, voltage drop, and load flow analysis), draft time current curves, and calculate raceway percent fill and seismic loading. In general, the validation and verification program was found to be adequate. However, the team found that the CAMS software used to calculate raceway percent fill and seismic loading had not been validated or verified prior to its use as required by Nuclear Engineering procedure. In response to this finding the licensee developed a Design Guide (DG-V.78) and completed the validation and benchmarking for the CAMS software features prior to the end of the inspection. The team reviewed the completed Design Guide and the benchmarking results of CAMS and found them to be adequate to verify all program features.

2.3.3 Fast Bus Transfer The team reviewed the fast bus transfer scheme which provided automatic transfer from the unit auxiliary transformer to the start-up transformer upon a reactor or turbine trip, opening of generator output breakers, manual operator transfer, or generator / transformer f ault . trip. The transfer control scheme was adequate, however, the team was concerned that the fast bus transfer scheme had not been analyzed for possible excessive dif ference in voltage between the auxiliary load bus and the incoming  ;

power source. Additionally, transfer time (cycles per second) has not been determined. These issues are current industry issues. The licensee ,

l was cognizant of the issues and related NRC generic communications and has demonstrated an approach to the resolution by authorizing EBASCO to perform a bus transfer dynamic simulation study. The licensee's demonstrated knowledge of the industry issues, and the proactive investigation and scheduling of a computer bus transfer analysis were considered a strength.

2.3.4 Cable Ampacity Ratings The licensee's calculation, RNP-E-5.004, " Analysis for Power Cable Ampacity Study," indicated certain safety related AC power cables were carrying load currents in excess of their evaluated ampacities. This was based on maximum ambient temperature (40 C) under LOCA conditions assuming  ;

cable trays contain over 43 conductors which requires maximum derating per l IPCEA P-46-426. This would result in a loss of cable life based on l exceeding the 75 C cable conductor temperature. The overloaded cables l were replaced. Based on the number of cases of marginal ampacity revealed '

by this analysis the licensee had initiated an engineering project PCN 88-192/04, to further evaluate and resolve the marginal cables and further analyze safety related 480 volt, 120/208 VAC, 120 VAC, and 125 VDC cables

1 8

l in the Reactor Auxiliary Building and the containment. The study was i limited to areas containing safety related cables. The team concluded the licensee's actions to resolve this !ssue were appropriate. No operability concerns were evident although this could be a factor in plant life extension evaluations.

l 2.3.5 Short Circuit Calculations l The team concluded that there is little margin for future EDS load growth based on short circuit calculations. The team reviewed computer program ASD0P calculation RNP-E-8.002, Rev. I for short circuit on 4.16 kV buses 1,2,3,and 4; 480 VAC dedicated shutdown busas; 480 VAC safety-related I buses El and E2; 480 VAC MCCs 5 and 6; and 208 VAC MCCs 9 and 10. The -

analyzed cases demonstrated adequate margin on the buses with no EDGs contributing to the short circuit. The licensee also analyzed the case with one EDG being exercised on either bus El or E2. The results of this calculation demonstrated bus ratings were exceeded. The licensee recalculated short circuits at El and E2 taking credit for internal circuit breaker impedances and less conservative voltage levels. The recalculation found bus El for symmetrical short circuit to be the worst case with a calculated value of 84.5 KA symmetrical and a rating of 85 KA symmetrical. This provides a margin of only .5 percent for future growth.

2.3.6 Voltage Drop Calculations The team found the calculations for postulated worst case loading to be adequate with acceptable margins. However, the demonstrated margins show that future load growth is limited.

The team reviewed AS00P calculation RNP-E-8.002, Rev.1 for voltage drop on non-safety 4.16 kV buses 1, 2, 3, and 4; 480 VAC dedicated shutdown DS bus; safety related buses El and E2 and safety related 480 VAC MCCs 5, 6, 16, and 18; and safety related 208 VAC MCCs 9 and 10. The minimum steady-state voltages with 115 kV switchyard toltages at minimum predicted value of 113.7 kV for buses El and E2 are calculated at 444 VAC and 446 VAC, respectively. Bus E1 with 444 Volts calculated and with a minimum require-ment of 429 volts had a positive margin of 3.5 percent. Bus E2 with 446 VAC calculated and minimum required of 428 volts had a positive margin of 4.2 percent. Safety-related 480 VAC MCCs 5,6,16, and 18 with calculated values 436, 439, 440, and 442, and with required minimum values of 420, 423, 419, and 419 have positive margins of 3.8%, 3.8%, 5.0%, and 5.5%,

respectively. The 208 VAC MCCs 9 and 10 with calculated values 189 and 188, and with minimum required 180 and 181 have positive margins of 5.0%

and 3.9%, respectively.

The minimum transient " ride-through" voltages for emergency buses considers minimum allowed motor terminal voltages while starting safety related motors during LOCA with the switchyard at 113.7 kV. Buses El and E2 with calculated voltage of 408 VAC and 410 VAC, and both with minimum required 362 VAC had margins of 12.7% and 13.3%, respectively. The 480 VAC MCCs 5, 6,16, and 18 with calculated values 399, 403, 384, and 385 t

r __

-4 i

9 and minimum required values 353, 356, 351, and 351 had positive margins of 13.0%,13.2%, 9.4% and 9.7%. The 208 VAC MCCs 9 and 10 with calculated values of 173 and 172 and with required values 158 and 159 had positive margins of 9.5% and 8.2 %.

2.3.7 Load Capacity Calculations The team reviewed ASDOP calculation RNP-E-8.002, Rev.1 for loading on the unit auxiliary and start up transformers; non-safety 4.16 kV buses 1, 2, 3, and 4; 480 VAC non-safety related buses 1, 2A, 28, 3, 4, and 5; 480 VAC ,

dedicated shutdown DS bus; safety related emergency 480 VAC buses El and E2; and MCC 9 and 10 480/208 VAC transformers. The team found loading to be adequate but noted safety-related buses El and E2 were marginal. Bus E1 with a 4000 A continuous rating and 3977 A calculated load had a .6%

margin. Bus E2 with 4000 A continuous rating and a 3923 A calculated had a 1.9% margin. The 4.16 kV buses 1, 2, 3, and 4 were marginal with calculated 3,114, 3,114, 2,816, and 2,816; and with maximum allowed currents of 3,257 A (value based on transformers being restrictive component associated with the bus); the margins were 4.4%, 4.4%, 13. 5%,

and 13.5%, respectively. The team considered the margins adequate for postulated worst case loading; however, as stated earlier for short circuit and voltage levels, there was little margin for future load growth. The team noted the licensee had initiated project PCN 88-084/100 to investigate possible alternatives to increase load, voltage, and short circuit margins for the entire electrical distribution system. The licensee had a load tracking program in place per Design Guide DG V.10 which required that all proposed load changes to the electrical dis-tribution system be evaluated, prior to implementation, to ensure these changes will not significantly reduce available margins. The team considered this program to be adequate.

l 2.3.8 Control Circuit Voltage Drop The team reviewed voltage drop in control circuit loop analysis for 4.16 kV switchgear control circuits, control circuits on El and E2 buses, control circuits for 480 VAC, and safety related control circuits for 208 VAC safety related motor control centers. The team found the analysis to be adequate.

2.3.9 Buses El and E2 Circuit Breaker Coordination The team reviewed circuit breaker coordination at the 480 VAC safety buses El and E2 and down to the connected MCCs and their loads. These calculations demonstrated coordination was acceptable.

2.3.10 Medium Voltage Protective Relaying The team reviewed the protective relaying schemes and the protective relay settings for the non safety related 4.16 kV switchgear and found. it satisfactory.

p

~

10 2.4 Emergency Diesel Generators L 2.4.1 Static and Dynamic Loading Analysis The team reviewed the licensee's calculation RNP-E-8.016, Static and i Dynamic Loading, which verified the appropriate sizing of the EDGs. The calculation evaluated the postulated worst case loading conditions, loss of coolant accident plus simultaneous loss of off-site power with only one EDG available, which embraced all other emergency operating conditions.

The calculation utilized electrical data provided either by equipment '

manufacturers or standard NEMA motor data. The EDG manufacturer (Colt) performed the dynamic loading analysis of the EDGs, using its own computer program. The team considered the methodology used for both calculations to be satisfactory and the conclusion appropriate. '

The dynamic analysis verified the minimum EDG output terminal voltage during the loading sequence was 81 percent of its nominal 480 V value and the maximum frequency drop was below 5 percent. These values were within the criteria provided in Appendix A of IEEE Standard 387-1984 and were  :

therefore considered satisfactory. Data from the transient analysis was used to verify that voltages on buses supplied from the EDGs did not drop below minimum cllowable levels during the EDG loading sequence. All bus voltages were found to remain within the acceptance criteria except for two 208 V buses. The licensee stated this was a transient dip and therefore motors on the buses would not stop operating; however, they were planning the installation of new transformers to rectify the problem.

The static analysis determined the maximum total load on EDG 'A' was 2500 KW and on EDG 'B' was 2528 KW. The maximum loads occurred at a time of forty (40) minutes following the LOCA/ LOOP event. The surveillance tests on the EDGs did not load EDG 'B' to the maximum value shown by the cal-culation. The licensee indicated that steps will be taken to modify the ,

tests and also to revise the UFSAR to reflect the load values shown in the I calculation. The static analysis verified that the EDG was capable of starting the largest motor when all other safety related loads were operating.

2.4.2 EDG Protection The team reviewed four recent calculations relating to EDG protection.

These covered: EDG ground overvoltage relay, EDG reverse power relay settings, EDG voltage permissive relay (27) set points and EDG neutral grounding resistor sizing. The calculations indicated that the existing protective devices meet the calculated requirements. The team considered the calculations to be technically accurate and well prepared. The licensee indicated that two additional calculations were planned to evaluate EDG ground overvoltage relay setpoints and EDG voltage controlled overcurrent relay setpoints.

I l

I

11 2.4.3 EDG Load Sequencing System The team reviewed the EDG load sequencing system and noted that an acceptable design was in place. New digital, solid state, class IE qualified Agastat timing relays had been installed. These provide greater accuracy and less drift than the original pneumatic relays. Interposing relays operating off of 120 VAC instrument buses were used to operate the motor feeder breakers.

2.5 Class IE Low Voltage AC Systems The team reviewed Class IE 120 VAC Vital Bus calculations and identified minor deficiencies. For example, the method of establishing channel loading relied on measurements rather than manufacturer's data. Secondly the minimum voltage at devices connected to the system had not been determined. The licensee stated that an instrument bus upgrade project was underway which will address these deficiencies and the overall configuration of the system.

The team reviewed the Instrument Bus Channel Loading Calculation, RNP-E-1.003, which established loading on each channel based on measurements taken over a 3 month period. This calculation verified that the loads on eacn main bus were conservatively below the rated values of the associated CVTs and inverters under all normal operating conditions.

The inverter on the h" train was restricted to a 5 kVA rating because of the limited capacity of the 125 VDC battery. The licensee demonstrated that loads under accident conditions would not change significantly from normal conditions.

The team reviewed a coordination study for the 120 VAC instrument buses calculation RN107-E-37-F, which verified protective device coordination was achieved, with two exceptions. These exceptions were of minor safety significance and related to the circuit breakers which feed buses IB7 and IB9 from IB2 and IB4 respectively. Circuit breakers to non-safety loads connected to IB7 and IB9 have the same rating (30A) as the feeder breakers. The team postulated that in the event of an earthquake, short circuits could occur in the non-safety loads and because of the non-coordination, the breakers supplying IB7 and 189 would trip thus completely de-energizing the buses. In response to the team's concern the licensee evaluated the safety loads supplied from IB7 and 189. The licensee was able to demonstrate that sufficient instrumentation would remain available to put the plant into a safe condition utilizing power from the other instrument buses. The licensee indicated that new 50A circuit breakers to supply instrument buses IB7 and IB9 would be installed. The lack of coordination is a identified as a team finding (see Appendix A, Finding 91-21-01).

The team reviewed a preliminary calculation, " Instrument Bus Voltage Evaluation" RNP-E-1.005, which determined whether bus voltages would remain within the established criteria when new instrumentation and control loads were added during refueling outage #13. The calculation

c - __

12 demonstrated that the bus voltages would remain above the minimum values provided that the loads on each channel did not exceed the stated CVT and inverter load limits. The given limits enveloped the existing loads as calculated in RNP-E-1.003 and the applicable calculation verified that when the new loads were taken into account, adequate margin still existed.

The team concluded that the system had adequate capacity based on the data presented.

2.6 Class IE 125 VDC System The team reviewed six calculations for the class IE 125 VDC system. These included: short circuit study, load profile and battery sizing, DC voltage profile, minimum inverter voltage verification and battery charger sizing. These calculations were evaluated against the requirements of IEEE Standards 485-1983 and 946-1985 and found to be acceptable.

The licensee indicated that future plans call for the re-habilitation of the 125 VDC system including the creation of separate battery rooms and the removal of the 125 VDC distribution gear from the battery room. The team considers that this would be a positive step. The team determined that the present system was adequate in terms of capacity and protection.

2.7 Containment Electrical Penetrations .

The team asked for assurance that the containment electrical penetrations were capable of: a) continuously carrying full load currents and b) withstanding short circuit currents. In response the licensee provided two hand calculations. One evaluated a 350 Hp motor (HVH fan) supplied from the class IE 480 VAC'switchgear and the other a 10 Hp motor supplied ,

from a 480 VAC MCC. The results of both calculations verified the ratings l of the penetrations were not exceeded. The team reviewed the calculations and found them to be technically accurate.

The team noted that the licensee's DBR project included a study that will assess all penetrations in terms of individual full load current and short circuit carrent capability. The study will be based on the original penetration manufacturer's data, the station load data and the bus short circuit levels. The team reviewed the scope of work document and a preliminary first copy of the assessment and concluded that the approach being followed was appropriate. The team also noted that the requirements for containment penetration leakage detection, and environmental and seismic qualification of the penetrations had been adequately addressed.

2.8 Grounding of AC Systems i

The Robinson plant utilized a high resistance grounding system which I allowed continued operation with a single ground fault. Ground detection relays provided annunciation but no automatic trip. The class IE systems' grounding will be analyzed as an element of the DBR project which o w ill l analyze the overall grounding system. This analysis will determine if the original design was adequate. The team evaluated the EDG neutral

13 grounding resistor and relay selection calculations and found them to be -

acceptable. A study prepared by the licensee in response to a question by the team indicated that the station ground protection schemes were <

adequate.

3.0 MECHANICAL SYSTEMS i

The team reviewed and evaluated the adequacy of mechanical systems ,

required to support EDS during normal operations and postulated accidents.

These systems included the EDG and EDG support systems, e.g. diesel fuel oil storage and transfer, air start, cooling water, lubrication oil, and the air exhaust and intake systems. Also reviewed was the plant service  ;

water interface with the EDGs. The team reviewed the HVAC for spaces i containing safety related electrical equipment to determine if equipment ambient conditions were maintained in accordance with vendor recommenda-  :

tions. Additionally, the anticipated accident mechanical loads which were used in the licensee analysis of the EDG loading were reviewed.

. Documents reviewed included applicable portions of the'UFSAR, engineering and vendor ducumentation, operating and maintenance procedures, mechanical ;

system calculations and , drawings, pump performance curves, equipment performance data sheets, and EDS related modification packages. The team >

verified that the licensee had documentation reviewed and approved by design to support the seismic qualification of EDS components. However, the adequacy of this documentation was not assessed as part of this inspection.

3.1 Conclusions i

The team concluded that the design and operation of the mechanical systems supporting the EDS were adequate. However, a number of items were found which merit further attention by the licensee. In the EDG accident load analysis, the containment fans may draw more power than stated in the calculation assumption which would reduce available loading margin. The ,

available service water flow to the EDG heat exchangers was less than recommended by the vendor. This could impact EDG capacity with high lake water temperature ( >95 degrees F) and EDG load greater than 2500 KW (100 to 110 percent load). ,

The licensee's analysis for ventilation of the E1/E2 safety related electrical equipment rooms under normal and accident conditions was found to be inadequate. The heat loads had not been analyzed to determine ambient temperatures and the subsequent impact on equipment. The licensee has initiated an analysis of this condition.

The licensee was unable to determine the status of the EDG underground fuel oil piping with respect to galvanic corrosion protection. The team  :

was concerned with the condition of this piping due to the possibility of corrosion induced reduction in pipe wall thickness which could affect the ,

seismic load capacity of the lines. The licensee initiated action to address this concern.

p -___

14 Performance of fuel oil sampling was not in accordance with industry standard practice related to ensuring a representative storage tank sample. The licensee initiated action to address this issue.

3.2 EDG Loading The team reviewed the licensee's postulated mechanical equipment loading values used in the EDG loading calculation. This consisted of reviewing equipment performance data and determining accident condition loads then translating these into electrical load values. The results were generally close to the values computed by the licensee.

3.3 EDG Support Systems The systems required to support the operation of the diesel generators were the jacket water, air starting, lubricating oil, combustion air ,

intake, combustion exhaust, and the instrumentation for the alarm and trip '

functions.

The team noted that lubricating oil pressures could be lower than the pressure in the jacket water system so that any leakage could be into the lubricating oil system which was opposed to normal industry practice.

Water forms emulsions which clog passages and could prevent proper lubrication, and was the cause of major repair work on EDG ' A' in the '

past.

The UFSAR indicates that the air start system for each diesel engine is sufficient for a 10 second attempt to start followed by 7 cold starts.

Although the team identified minor inaccuracies in the verifying docu-mentation, the air start capacity was essentially as stated.

3.4 Plant Service Water Interface The service water flow for the EDG heat exchangers was. less than the manufacturer recommendations. The licensee had previously identified this condition and developed evaluations to address this issue concluding that the available flow was adequate for all postulated conditions. While the team agreed that adequate service water flow existed for most conditions, there was uncertainty regarding the adequacy of flow d" ring high lake water temperatures and EDG load greater than 2500 KW. A recent transient analysis (discussed in paragraph 2.4.1) indicated the possibility of EDG loading greater than 2500 KW during certain accident scenarios.

The EOG manufacturer's flow recommendation for 95 degrees F service water (SW) was 600 gpm for operation at 100% electrical power (2500 KW output) and 700 gpm for 11D% electrical power (2750 KW output). SW verification tests have shown that the SW flow rate to each EDG following a LOCA would

' be in the order of 550 gpm. A number of analyses were performed to verify the functional capability of the EDG's with the reduced SW flow rates.

15 The analyses submitted to the team for inspection were based on a variety of assumptions, some of which were conservative in the context of ensuring adequate cooling flow, and others which were not. All analyses used the data from a manufacturer's drawing specifying the EDG heat exchanger conditions and performance, (Dwg 11 905 210) but for which the supporting manufacturer's thermal analysis was apparently not locatable. Essential information for the use of this drawing include its purpose i.e. design, testing, installation, mid-life performance, etc. each of which would require different margins on the parameters and the test conditions under which the measurements were made. Furthermore, the actual heat loads to be dissipated by the heat exchangers depend on specifics of the diesel installation and any modifications. The analyses assumed the design drawing information was the result of manufacturer testing. This assumption was inaccurate. The manufacturer could not verify that a 110 percent load test had been performed. This test would have supported the licensee's conclusion regarding adequate service water flow for EDG operation greater than 2500 KW.

The n.anuf acturer's representative also said that fouling factor information should have been communicated to the heat exchanger fabricator along with Drawing 11 905 210, but doubted that Colt would have retained any formal record of that transfe . The fabricator, ITT Standard, could not be contacted for more info.me , ion. Actual installed heat exchangers' performance data was not used it the analyses.

The licensee's analysis did not verify that service water flow was

- sufficient to support EDG loading between 100 and 110 percent load when heat sink temperature was 95 degrees F. Predictions for cases with conditions close to those expected following LOCA's indicate LO and JW temperatures will be near or above alarm thresholds. Operation in this range permits very little margin for error. This margin may have been eroded by inaccuracies in the design i n fo rmati on . These inaccuracies challenge the accuracy of the licensee's overall conclusion (see Appendix A, Findino 91-21-02).

3.5 Heating, Ventilation, and Air Conditioning The team reviewed HVAC design for the EDG rooms and other safety related equipment spaces to ensure ambient conditions were maintained within equipment design requirements. With the exception of the E1/E2 equipment room, the HVAC was adequately designed.

The original HVAC design for the E1/E2 room was based on contamination control rather the ambient temperature. The redundant buses powering vital equipment, as well as Reactor Protection and Safeguards Logic Cabinets, were located in this Auxillary Building room. The licensee had not evaluated ambient conditions in this space and potential impact on safety related equipment due to rising room temperatures t esulting from loss of ventilation. Room ventilation is lost during a loss of offsite power.

mi g eemWheAJ + A m J's e' A.a R

16 Following identification of this issue by the team, the licensee initiated an analysis of the room heat loads and potential ambient conditions during postulated accident conditions. The analysis provided by the licensee during the inspection was not fully conclusive. Although no operability issue was evident the issue required further evaluation. The licensee stated testing to accurately determine heat loads would be accomplished and the results evaluated to determine if further action is required (see Appendix A, Findino 91-21-03).

3.6 EDG Fuel Oil Storage and Transfer System 3.6.1 Inventory and Fuel Consumption The fuel oil system consisted of a diesel fuel oil storage tank, two transfer pumps, two independent underground FO transfer lines to each day tank in each EDG room, piping from the day tank to the engines and '

connecting underground piping between tanks.

Operating procedures required a minimum F0 storage of 21,000 gallons in the DFOST and 15,000 gallons in the I-C Turbine F0 tank. The capacities of those tanks was ' o 'O gallons in the former and 95,000 gallons in the latter. With a low level alarm at 20,000 gallons, and assurance that fuel inventories in the I-C Turbine FO tank were always well above the required minimum 15,000 gallons, a minimum total inventory of 35,000 ,

gallons was available. The team concluded that fuel storage requirements i were met.

3.6.2 Miscellaneous Fuel Oil Issues  !

The team identified two issues related to the diesel fuel oil. The first j issue was the indeterminate condition of the underground fuel oil piping due to malfunction of the system which provided galvanic corrosion protection for this piping. A cathodic protection system was installed in 1982 which began to malfunction in 1988 and was discontinued in 1990.  !

The licensee was unable to determine what methodology was provided for galvanic corrosion protection on the original piping installation.

Additionally the licensee could not verify the integrity of the piping.

Galvanic corrosion could result in pipe wall thinning which would impact the seismic irtegrity of the piping. The licensee had already initiated actions to evaluate the Cathodic Protection System functionality. The licensee further stated that examination of a portion of the piping had been scheduled for 1992 (see Appendix A, Finding 91-21-04).

The second issue was related to the licensee's methodology for sampling of the EDG fuel oil storage tanks. The team concluded that the methodology used did not result in a representative sample of the tank contents. The applicable industry standard requires sampling at various levels within the tank. The licensee sample is drawn from a location 15 inches in from  !

the tank side and 15 inches above the tank bottom. The licensee stated that their methodology was adequate based on a history of samples from i

this location being within the acceptance criteria. The team disagreed

~ - ---

u__.._.

17 that this basis justified deviation from industry practice to which the licensee was committed. The licensee initiated actions to acquire sampling equipment which would more accurately determine if the current method is equivalent to industry guidance (see Appendix A, Finding 91-21-05).

4.0 MAINTENANCE, TESTING, CALIBRATION, AND CONFIGURATION CONTROL The team performed walkdown inspections of the EDS to identify the material condition of the electrical equipment and panels. Portions of the "as installed" configuration of the EDS were examined to determine its compliance with design drawings and documents. The electrical maintenance program, procedures, surveillances, and work orders were reviewed to ensure the EDS was being properly maintained to function for the life of the plant. Data sheets from completed calibration and surveillance procedures were reviewed to verify the EDS operates in accordance with design specifications and requirements. The method used for fuse control configuration was examined to ensure correct sizes and types were installed. Preliminary relay setting drawings were reviewed to determine if the licensee was in the process of developing and implementing an effective program for controlling setpoints for protective relays, over-load relays, circuit breakers, and switchgear. Testing and surveillance procedures for the emergency diesels generators were reviewed to determine if specifications are being met.

4.1 Conclusions In the area of configuration control the team found that the Nuclear Engineering Department was in the process of upgrading the electrical drawings; developing new one-line drawings for the 480 V Class IE MCCs; developing " relay setting" drawings; and implementing a fuse control program.

The overall material condition of the EDS was found satisfactcry except for spare electrical leads. The ends of spare cables, wires, leads, and conductors were not always securely taped (capped) to last for the life of the plant. Additionally, examples were identified of improperly labeled electrical panels. The team identified that the Class IE motor starters for two safety related RHR MOVs were undersized. The team identified examples of potential undetectable failures in plant MOVs. The licensee initiated an investigation of this condition.

In the area of maintenance the team noted the licensee was still using i check lists for preventive maintenance procedures. This was identified as a weakness during the NRC Maintenance Team Inspection. However, the team verified the switchgear was being maintained in a PM program and the protective relays were calibrated on a scheduled basis.

In the area of testing and surveillance the team identified that the EDGs had not been tested at nameplate rating. This was identified as an unresolved item. In addition the EDG's surveillance procedures do not l

18 require load acceptance and/or rejection be verified. In another area, the team identified that DC ground indication trending had not been performed.

4.2 Equipment Walkdowns The electrical components examined during inspection walkdowns included fuses, protective relays, motor starters, circuit breakers, switchgear, batteries, chargers, inverters, cables, cable trays, transformers, panels, and cubicles. The associated components, equipment, and panels in the following electrical systems and areas were inspected:

y The safety related 125 VDC system, batteries, chargers, power panels, ,

and 120 VAC inverters. '

The safety related 480 V buses, switchgear, protective relays, and motor control centers.

The 4.16 kV switchgear, cubicles, panels, and load centers.

The 480 V emergency diesel generators and control panels.

The main step-up, the unit auxiliary, and the startup transformers.

The 230/115 kV main switchyard.

. Walkdown inspections were conducted by the team to determine the EDS conformance to design requirements. Design drawings used for the field inspections reflected the "as-installed" configuration. Specific atten- I tion was directed to the Class 1E MCCs since modifications were recently installed. The team examined the f uses, moter starters, transformers, circuit breakers, and cubicles in three different MCCs. The material condition of the electrical panels, MCCs, switchgear cubicles, and relay  ;

panels was satisfactory except for cables arid wires and MCC compartment  ;

labels. 1 The ends of some spare cables, wires, leads, and conductors were not securely taped to last for the life of the plant. In addition, the team observed several examples in which MCC compartment labels did not agree .

with descriptions on the electrical line-up - procedures. The licensee  !

initiated a project to upgrade identification of plant equipment in the '

first quarter of 1991. The team identified the spare cable terminations and MCC label issues as poor maintenance practices (see Appendix A, Finding 91-21-06).

During the inspection of safety related MCCs E and 6, the team identified that the motor starters in compartments IJ and 12J respectively did not  !

appear to be sized to match the circuit breakers. The licensee confirmed l that the motor starters for RHR valves 744A and 744B were undersized ~ and would be replaced with the correct size (see Apoendix A. Finding 91-21-07). The team reviewed the wiring drawings for these valves with i

i l

I 19 l

l licensee's engineering personnel and determined that no method for identifying a tripped overload relay was available to the plant operators.

A tripped overload relay in a motor starter circuit would prevent the valve from operating. The team identified this condition as a potential undetectable failure for the MOVs (see Appendix A, Findino 91-21-08).

The electrical drawings used during the walkdowns had several minor errors such as the location of components. The licensee agreed the drawings '

needed to be upgraded and stated an upgrade program was in process. The <

licensee was also in the process of developing one-line drawings for the 480 V MCCs. The team reviewed these preliminary drawings and considered them to be appropriate, particularly in support of maintenance activities.

4.3 Equipment Maintenance, Testing, and Calibration The team inspected the maintenance program to ensure that the EDS was being properly maintained to function for the life of the plant. The calibrations, testing, ;surveillances, and completed work orders were reviewed to determine if the EDS was operating within design requirements and technical specifications. Maintenance was performed by two groups.

The high voltage-(230/115 kV) equipment and main switchyard were serviced by an offsite " Transmission Group." The Electrical / Maintenance Department on site performed all other maintenance activities for the low voltage equipment and EDGs.

4.3.1 Preventive and Predictive Maintenance In early 1991, the licensee initiated a project to review preventive maintenance (PM), predictive, and monitoring activities necessary for safety related and balance of plant systems. The review had been completed for the 480 and 120/208 VAC and 125 VDC systems. The EDG review was in progress. Resolution of the recommended changes / additions to the PM program were scheduled for completion by December 31, 1993. Examples of vendor recommendations not currently contained in PM procedures include failure to verify generator brush spring tension and measurement of ring eccentricity as described in the vendor instruction manual no. 3073B, dated March 1960. These specific examples and other examples were identified by the licensee's PM review project.

4.3.2 MCCs, Protective Relays, and Switchgear The team reviewed the maintenance and calibration programs for both the Transmission Group and the Maintenance Department. PM and calibration procedures were reviewed to determine content and details. The Transmis-sion Group had completed calibration cards verifying the protective relays for the 230/115 kV switchyard were calibrated and maintained in the PM program. The Maintenance Department PM procedures used brief " check lists" to identify required work. The use of these check lists was identified as a weakness by the NRC Maintenance Team in Inspection Report 50-261/90-10 dated August 9, 1990.

w 20 1

The PM schedule and completed work orders were reviewed to verify that the 480 V and the 4.16 kV switchgear was maintained on scheduled basis.

Completed calibration data sheets and relay settings were reviewed to verify the protective relays had been calibrated. The material condition of the switchgear indicated regular maintenance was satisfactorily performed. The material condition of the motor control centers was good.

The NED was in the process of developing a fuse control program. The licensee stated the program for Class 1E application would be comple'.ed

, for the next refueling outage (Spring 1992). The team reviewed the l preliminary contents and concluded the fuse control program described would be sufficient for ensuring the correct size and type of fuses are specified. Overall the team detemined that NED had good programs and engineers to support maintenance activities.

4.3.3 EDG Surveillance The team reviewed EDG surveillance test activities to determine if the

, EDGs met design and TS requirements. During the review of completed surveillance tests, the team identified that the EDGs had not been tested at nameplate rating as required by TS 4.6.1.1. Both the slow speed and rapid speed start surveillance tests. OST-401 and 409, required the EDGs to be loaded to 2500 KW. The TS requirement was for assumption of load up to the nameplate rating which is 2500 KW at 80 percent power factor and 3125 kVA. Based on this information the team concluded the licensee has apparently not been testing the EDGs as required by TS (see Appendix A, j

. Findino 91-21-09).

Review of startup and factory test reports demonstrated that the generator had not been tested to 3125 kVA. In May of 1987, special test SP-776 and SP-777 was performed to measure the A and B EDGs governors' character-istics. During these tests the A and B EDGs were tested to 3014 kVA and 3172 kVA respectively which enveloped the nameplate rating only for the BEDG but enveloped both the A and B EDGs anticipated maximum accident kVA loading of 2880 kVA and 2886 kVA (calculation RNP.E-8.002 revision 1).

Since neither OST-401, nor 409 functionally test the EDGs to anticipated maximum accident kVA loading, the functional capability of the EDGs to provide accident kVA loads had not been demonstrated since the special tests discussed above. Vendor information indicated that the generators were capable of supplying several hundred additional kVA without modifi-cations. The licensee stated that the appropriate DSTs would be revised to test the EDGs to a kVA equivalent to the nameplate rating. The testing would be implemented by December 9, 1991.

The team also noted that the present test procedures did not perform other testing such as load acceptance and rejection tests and extended time run tests which are routinely performed in the industry. Generally, such testing is required by TS or commitment to Regulatory Guide 1.108; however, HB Robinson was not committed to this Regulatory Guide nor was such testing required by the TS. The licensee stated their intention to review the value of such testing for possible incorporation into their

= _ . .

I

. t l

I 21 testing program. Additionally, as discussed in section 2.4.1, the EDG was l not periodically tested to verify performance above 2500 KW which the j static load analysis, discussed in paragraph 2.4.1, demonstrated was a  :

possible accident load condition. {

i On September 14, 1991, the licensee identified that established i surveillance test procedures had not fully tested each loss of voltage l load shed channel (TS Table 3.5-3, item 3.a). They also identified that j testing had not completely demonstrated the load shed function upon a 1 simulated loss of all normal AC coincident with a safety injection signal ,

(TS 4.6.1.2). Additional information concerning this item and the ,

associated Waivers of Compliance which allowed continued operation is l contained in NRC Inspection Report No. 50-261/91-20. The team reviewed -

the loss of voltage, the degraded voltage and load sequencing test  ;

procedures. Test deficiencies other than those already identified by the licensee were not identified.

5.0 ENGINEERING AND TECHNICAL SUPPORT  ;

The team assessed the licensee's capability and performance regarding .:

engineering and technical support associated with the EDS. The basis for ,

this assessment included the following areas; technical organizations and interfaces, problem identification and resolution, support of routine EDS related activities, and modifications. >

5.1 Conclusions  :

. I Engineering and technical support for EDS related activities and design controls for EDS systems and components were adequate to monitnr and '

maintain the functional integrity of the EDS as designed. Engineering -

organizations were adequately staffed to provide EDS support. Responsi-bilities and interfaces for engineering organizations were appropriately defined and documented. The team concluded the engineering and technical support organizations demonstrated a good knowledge level regarding the l EDS and associated components. Documentation of plant activities demon-strated engineering involvement in problem identification and resolution as well as routine plant activities. Appropriate design controls were implemented on EDS related modifications.

l 5.2 Organization and Staff  !

I Engineering and technical support for EDS related activities was provided  :

by onsite and offsite engineering organizations. The Nuclear Engineering  !

Department was designated as the design authority and was located  !

primarily offsite with a smaller onsite branch staff. Technical Support -!

was the onsite technical authority and included System Engineering as the i largest group. System engineers provided the focal point for technical '

support of EDS related activities, interfacing with the plant staff, NED, and vendor technical representatives. The Technical Support staff of i approximately 70 personnel included 35 percent contractors. Review of [

I

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

paa -

22 plant activities during this inspection indicated that staff size was adequate to provide required technical support for the EDS .

Responsibilities for engineering groups were appropriately defined and documented in plant memorandum and guidance manuals. General knowledge of the Technical Support and NED staffs was good. This was demonstrated by staff response to team inquiries regarding systems' and components' design, configuration, testing, and maintenance. Additionally, the team observed that interfaces were effective between NED and Technical Support.

Overall, adequate programs were developed to provide technical support of EDS activities. Performance by the engineering organizations indicated appropriate implementation of these programs.

5.3 Problem Identification and Resolution The team assessed engineering and technical support involvement in problem identification and resolution activities. Involvement was assessed by review of deficiency reporting programs and trending activities related to the EDS. Deficiency reporting documentation from the previous five year period was reviewed. Resolution of identified deficiencies was adequate.

Cause analysis for deficiencies was generally adequate. An exception was FRs which programmatically did not require cause analysis, and inconsis-tent cause analysis on corrective maintenance work requests. An upgrade of the corrective action program in December, 1990 has improved cause analysis activity for deficiency reporting processes.

. Several examples demonstrated engineering involvement in problem identification and resolution. SCRs88-015 and 89-002 identified water in EDG cylinders. Extensive investigation activities demonstrated technical staff involvement in eventual resolution of the problem. SCR 90-020 identified a problem associated with the Auxiliary Building smoke alarms being activated by the exhaust smoke generated by an EDG fast start. The technical evaluation of this condition was detailed and comprehensive. A PIR dated June 13,1990 was initiated by a System Engineer to identify the malfunction of the cathodic protection system for the underground fuel oil piping. POER 88-02 addressed a series of overspeed trips on both EDGs in 1987 and 1988. A comprehensive evaluation of EDG design and operating characteristics by Technical Support resulted in corrective maintenance which eliminated the problem. An additional indication of involvement in problem identification was the large percentage of ACRs which were initiated by Technical Support organizations in 1991.

The team reviewed trending activities performed by the engineering staff. l The system engineers accomplish trending for assigned systems. Fuel oil pumps and valves are trended by the ISI program. High EDG exhaust cylinder temperatures were identified as a result of trending and was <

under investigation. An apparent decrease in fuel oil transfer pump capability was also identified which was attributed to inappropriate )

testing methodology. In general, review of deficiency reporting processes and trending activities demonstrated appropriate involvement of the i

i I

y _ __

23 engineering and technical support organizations in problem identification and resolution.

5.4 Routine Plant Activities Engineering and technical support involvement in routine plant activities was reviewed by the team. Routine involvement included maintenance, testing, and operations activities. Additional routine involvement was provided by an EDG system team which routinely reviews EDG performance and activities. An electrical systems team was developing preventive main-tenance procedures for the generator component of the EDG. Review of maintenance documentation demonstrated frequent requests for engineering assistance in accomplishing maintenance tasks. Preventive maintenance, which included periodic disassembly of EDG components for inspection and testing, was supported by the licensee's technical staff and vendor technical representatives during performance. Surveillance testing of the EDGs demonstrated interface of operations, maintenance, and technical staffs in recording and evaluation of performance information. Overall, review of maintenance, testing and operational documentation demonstrated appropriate engineering and technical support involvement in routine activities.

5.5 Modifications The teams review of EDS related modifications included electrical and mechanical modifications developed and implemented within the previous 10 year period. Temporary modification activity was also reviewed. The following modifications were reviewed:

MOD 955 EDG Upgrades MOD 763 Add New Pressure Taps for Controlling EDG Air Compressors MOD 921 EDG Room Ventilation Louvers Modifications MOD 922 Remove Trip Function from MCC 5 Transfer Switch Breakers MOD 939 Protective Equipment Upgrade MCC 5,6,9,10 Safety Evaluations were appropriately detailed and post modification ,

testing was adequate. Overall, design development and implementation was <

adequate. An exception was MOD 939 in which relay trip setpoints for MCCs I were inadequate to preclude spurious trips. This error was identified l during post modification testing and a comprehensive root cause analysis i was performed. The cause was identified to be an inaccurate design  !

document referenced to determine setpoints. Documentation of modifica-  !

tions had improved over the time period reviewed. A good practice noted '

in the more recent modifications (MDD 955) was the designation of organizational responsibilities for various aspects of the design change within the modification package.

The team reviewed temporary modification controls and available examples of temporary modifications related to the EDS. The licensee's procedures  !

g __

24 provided adeqeate controls for temporary modifications. Temporary Modification 91-718 installed an electrical jumper around "A" station battery cell 54 This was the only EDS related temporary modification.

Appropriate design controls were implemented. A temporary modification to provide cooling water to the EDGs for maintenance runs during the outage was accomplished via a special procedure, SP-947 dated November 19, 1990.

Design controls were adequate for control of this temporary modification.

The licensee's approved design control program permitted use of special procedures for implementation of this temporary modification. Based on the sample of modifications reviewed, adequate design controls were implemented for EDS related modification activities.

6.0 EXIT MEETING The team met with the licensee's representatives (denoted in Appendix C) at the conclusion of the inspection on October 25, 1991, at the plant site. The findings were discussed at that time. No dissenting comments were received at that time however the licensee's management met with the Senior Resident Inspector on November 1, 1991 to dissent with the finding related to inadequate surveillance testing of the EDG, The licensee's stated position was that the applicable operational surveillance test procedures, OST 401 and 409, tested the EDGs as required by TS. The Senior Resident Inspector observed an attempt to test the A EDG to nameplate rating on December 9,1991. The test was accomplished, but caused high voltages on the equipment due to lack of load on the El bus.

Due to plant configuration, voltage will be even higher on the E2 bus, therefore that test was suspended pending further study of how it should be performed without equipment damage.

NRC management has determined this item will be designated as unresolved pending further review by the NRC to determine the true intent of the TS surveillance requirement. Licensee representatives were informed of this decision just prior to report completion. Proprietary information is not contained in this report.

l I

l

a APPENDIX A FINDINGS FINDING 91-21-01: Inadequate Coordination Between Safety and Non-Safety Circuit Breakers on the 120 VAC Vital Bus System (para. 2.5)

DESCRIPTION:

The team noted that the supplies to instrument buses IB7 and IB9, from instrument buses 182 and IB4 respectively, were through 30A circuit breakers and that non-safety loads connected to IB7 and IB9 also utilized 30A breakers.

The team postulated that in the event of an earthquake both non-safety loads could fail short-circuit and because of lack of coordination the feeder breakers may trip, thus completely de-energizing both IB7 and IB9.

The design calculation RN107-E-37-F, " Coordination Study for Instrument Buses,"

issued in 1986 had recognized the lack of coordination and had recommended changing the feeder breakers to new ones with a 50A rating with a corresponding increase in cable size. These recommendations had not been implemented at the time of the inspection. In response to the team's concern the licersee conducted an investigation into the effects on the station if the above situation developed. The licensee was able to demonstrate that the plant could be shutdown safely and that the auxiliary systems required for heat removal would remain available with instrumentation power provided by the remaining buses.

The team accepted the licensee's position but was concerned that the above scenario had not been recognized by the licensee.

The team noted that the licensee had initiated a long term " Instrument Bus l Upgrade" project (PCN 85 - 032/04) which proposes hardware changes and a possible system re-configuration under which the present problem would have been corrected. However the licensee indicated that they would not wait for the results of this project but would initiate action to purchase new 50A circuit breakers.

SAFETY SIGNIFICANCE:

A maximum hypothetical earthquake could result in the failure of redundant I safety buses.  !

y ._

4 Appendix A 2 Finding 91-21-02: Marginal Service Water Flow Rate to Diesels for 110% Power (para. 3.4)

This finding is identified as an unresolved item.

DESCRIPTION:

Service water flow rates to the diesels for cooling under LOCA conditions have been shown by tests to be below the manufacturer's recommendation. While the manufacturer has documented a need for 600 gpm of 95 degrees F service water at 100% power operation and 700 gpm of 95 degrees service water at 110% power, the special SW tests showed that the maximum available flow to be expected under LOCA conditions would be in the order of 550 gpm.

The question of whether the lower SW flow rates are acceptable hinges on the performance of the heat exchangers. As no in-service tests of the heat exchangers are available (Reference 03-1), calculations were performed to

' ascertain whether the measured SW flow rates were sufficient under LOCA conditions.

A summary of the status of the calculations was given in the DBD validation report. The team disagreed with the validation conclusion that adequate SW is available for 110% power. Actual installed heat exchangers' performance data was not used in the analysis of low service water flow conditions. The analyses were based on ideal conditions related to a factory tested newly assembled EDG which may not be accurate for the present installed condition.

The conclusion that 550 gpm flow was adequate permits operation of EDG support systems in the alarm range at greater than the 100 percent EDG continuous rating of 2500 kw. In the alarm range the margin for error is limited before damage to the EDG could occur with subsequent impact on the EDG safety function. Calculation non-conservative assumptions discussed above may have consumed the available safety margin before instrument errors have been accounted for. Additionally, a recent licensee EDG loading analysis indicated possible loading above the continuous loading value. Based on the potential for greater than 100 percent continuous rating load in conjunction with high lake temperatures, the team concluded this specific scenario required more accurate analysis to verify the EDG could meet its required safety function under these conditions.

SAFETY SIGNIFICANCE:

A recently completed transient analysis by the licensee indicated that in certain scenarios EDG may be required to provide between 100 and 110 percent load. The licensee's analysis did not verify that service water flow was sufficient to support this EDG loading when heat sink temperature was 95 degrees F.

U

pa---- .- 1 Appendix A 3 Finding 91-21-03: E1/E2 Equipment Room Ambient Conditions Not Evaluated (para. 3.5)

This finding is identified as an unresolved item. l 1

DESCRIPTION:

The redundant 480 VAC safety-related buses, El and E2, are located.in the same room in the auxiliary building. Equipment in this room included safety-related equipment breakers and inverters for vital control power and Reactor Protection and Safeguards Logic Cabinets. The ambient temperatures resulting from electrical equipment heat loads had not been analyzed to determine potential impact on equipment performance or if temperatures remained within equipment design requirements in normal and accident conditions. Ventilation in this room is non-safety, i.e. lost on a loss of offsite power.

Following NRC questions, the licensee initiated an evaluation of heat loads in this space for normal and abnormal conditions. An analysis based on industry reference values for equipment heat loads was performed. Additionally, an analysis based on informal testing accomplished in 1985 was performed. The results of these analyses varied considerably. The former indicated an ambient temperature of 136 degrees F in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, leveling at 166 degrees in 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

The latter analysis indicated 126 degrees F in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, leveling at 144 degrees in 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The difference in the computed heat loads was the cause of the variation. The industry reference heat loads were 52 KW, the test based value was 17 KW.

The team concluded that, although no operability issue was evident, further analysis was requi red to more accurately determine the heat loads and subsequent ambient temperatures. This was based on two factors: 1) the large variation in computed heat loads, and 2) the test was an uncontrolled, informal activity which required several assumptions to be made regarding actual ventilation flow and accuracy of instrumentation used. The license stated during the inspection that a controlled test would be performed and evaluation of equipment impact accomplished by the end of 1992.

SAFETY SIGNIFICANCE:

Electrical equipment performance could be impacted when exposed to high ambient temperatures. This could require derating or other compensatory measures to assure adequate equipment functionality.

y _ _ . . , . .

i I

Appendix A 4 Finding 91-21-04: Corrosion Protection of Underground Fuel Oil Piping (para.

3.6.2).

DESCRIPTION:

The cathodic protection system was installed in 1981 to prevent galvanic corrosion from piping to ground. However, the system is known to have been operating outside of its original specification since August 1988 and thus the duration of. protection has only been about 7 years. The Licensee was unable to provide documentation which specified what provisions for galvanic corrosion had been made at the time of .the original installation. As a result, the licensee was unable to determine the present condition. of_ the underground fuel piping. Degradation of the cathodic protection system in'1988 appeared to have been caused by installation of concrete in the yard.

The licensee stated that a sample of the underground fuel oil piping would be inspected during the 1992 refueling outage: Further action will be based on inspection results. The licensee scheduled a technical representative to t review the onsite cathodic protection system and upgrade the site staff ,

knowledge of the system. ~

SAFETY SIGNIFICANCE: l If corroded the underground piping could be susceptible to failure ~ during a '

seismic event which could result in loss of fuel oil between the storage tanks +

and the EDG day tanks.

~

I:

/

i Appendix A 5 t

Findino 91-21-05: Fuel Oil Sampling Methodology (para. 3.6.2)

This finding will be identified as a deviation.

DESCRIPTION:

Sampling of the fuel oil in the diesel fuel oil storage tank (DFOST) was-performed at a single location near the bottem of the tank. The .irrdustry standard methodology for sampling, to which the licensee is committed, ensures that the sample is representative of the tank contents by sampling at different ,

tank elevations or by circulating tank contents prior to sampling.

Deviation from the methods described in the industry standard, ASTM D270, are l acceptable with appropriate justification. No justification for this deviation, either comparative testing or other demonstration of equivalency, was documented.

TECHNICAL REQUIREMENTS:

Regulatory Guide 1.137, C.2, states that ASTM D270 should be used to sample fuel oil in tanks and'the purpose was to obtain a representative sample of tank contents.

SAFETY SIGNIFICANCE:

Failure to ensure representative testing of the diesel fuel could cause

. degradation of fuel oil to go unnoticed and lead to a failure of the diesel generators in an emergency situation.

5 t

F m --____.-___m_____

y _

Appendix A 6 FINDING 91-21-06: Examples of Poor Maintenance Practices (para 4.2)

DESCRIPTION:

The team conducted walkdown inspections of the electrical panels, motor control centers, switchgear , and relay racks to determine the material condition of the equipment and the wiring. During these walkdowns the following deficiencies were identified. A number of spare electrical cables, wires, leads, and conductors ends were not securely taped to last for the life of the plant. In several instances the tape appeared to be loose. The functional description on MCC compartment labels may vary substantially from that contained in OP-603 and in some cases are technically incorrect.

The licensee stated the standard practice for terminating spare cables and conductors at the site has been to wrap the ends with tape. The licensee agreed with the team's finding that some spare cables and conductors ends are not up to standards. As a result of this finding the licensee initiated ACR 91-370. ACR 91-370 required the practice of terminating spare wires be investigated and resolution be determined. In addition Work Requests WR/JO 91-APHG1, 91-APHH1, 91-APHII, 91-APHJ1, and APHK1 all dated October 23, 1991, had been written for corrective action.

The functional description on MCC compartment labels may vary substantially from that contained in OP-603 and in some cases are technically incorrect. An example of the former was MCC compartment label which read " SAT SUCTION VLV SI-845B" whereas OP-603 referred to. this compartment as " SPRAY ADDITIVE TANK

. OUTLET ISOL. 51-8458." An example of the latter involved a spelling error on a MCC label (section misspelled as suction) resulting in an MOV compartment label of " AUX. FWP SUCTION VA. V2-20A." The OP-603 reference, " Motor Driven AFW Pump Discharge Cross-Connect V2-20A," was technically correct. The licensee initiated a project to upgrade identification of plant equipment in the first quarter of 1991. This project will include mechanical components such as valves as well as electrical components such as MCC and electrical distribution panel breaker labels. However, the licensee has not determined the style, type or means of attachment of labels to be used nor developed a schedule for the labeling of electrical components.

SAFETY SIGNIFICANCE:

Spare dangling electrical conductors could potentially cause a short resulting in the failure of a safety system. In addition these conductors could  ;

potentially shock plant personnel if voltage was present. Incorrect or '

inconsistent labeling could result in improper component isolation for maintenance or maintenance on incorrect equipment.

l

)

i

p _x, .

- -- - --= ---- ' - -

o' Appendix A 7

]

1 FINDING 91-21-07: The Motor Starters For Motor Operated RHR Valves 744A And 7448 Are Undersized (para. 4.2)

DESCRIPTION:

The team identified that in MCC 5 Compartment IJ and MCC 6 Compartment 12J 50 ampere motor protector. circuit breakers were feeding Size 1 motor starters.

The licensee was requested to verify that this configuration was correct. The -

licensee's engineering staff performed an Operability Determination 91-022 to '

analyze this condition. The licensee determined the motor starters should be a i Size 2. The licensee analyzed the affect of having undersized motor starter in i the circuits and concluded that the valves were operable. However, the licensee stated the motor starters will be replaced with Size 2 starters no later than refueling outage No. 14. Until then, the existing motor starters are to be inspected each time the valves are cycled. The team agreed with the licensee's analysis and proposed corrective action. i SAFETY SIGNIFICANCE:

A failure of the motor starters would prevent the motor operated RHR Valves  ;

744A and/or 744B from cycling which could compromise the Low Head Safety Injection flow path. .

4

l i

)

l i

1 i

I 4

, , , . . - , , , _ __ - _m, . ~ . . .

__s * .- -

- - =- - -- - ^ - ' - - - - -

o 4

l Appendix A 8 FINDING 91-21-08: Undetectable Failure M'echanism on MOVs (para. 4.2) i DESCRIPTION:

The team identified that there is no indication provided to plant operators to detect a tripped overload relay in a motor starter circuit. This was  ;

identified by the team during the review of the motor starters for MOV 744A and '

MOV 7448. The power to the position indication lights (OPEN and CLOSED) is fed -

directly from the fuse side of the circuit. It is not interlocked to the overload relay and therefore the position lights can not be used. The licensee agreed with the team that a tripped overload relay could not be identified.

The licensee was requested to address this concern by the team. The licensee l stated that an adverse condition report (ACR) would be issued to investigate

• this concern. However, the licensee stated that the MOV circuits were of original plant design and did not represent an. operability concern. The team considered this finding as a potential " undetectable failure."

SAFETY SIGNIFICANCE:

A tripped cierload relay would not be detected and the valve would be inoperable cee to a loss of electrical power.

1 i

4 d

- Appendix A 9 f

Finding 91-21-09: EDGs Not Tested at Name Plate Rating as Required by TS 4.6.1.1 (para. 4.3.3)

This finding is identified as an unresolved item, Failure to meet TS require-ments'for EDG testing.

DESCRIPTION:

TS 4.6.1.1 requires operation of the EDGs with an assumption of load up to the nameplate rating. The EDG's nameplate specified a rating of 2500 KW at- 80 percent power factor and 3125 kVA. OST 401 and 409 which implemented this TS requirement tested the EDGs at 2500 KW at an unspecified power factor or kVA loading. Subsequently, the licensee attempted tests at 3125kVA and found this caused very high voltages on equipment due to light loading on the E buses. It is not the NRCs intention to cause the licensee to push test conditions to the point of causing equipment damage. Therefore the correct intention of the TS surveillance requirement needs further consideration by the NRC and CP&L. This item will be considered unresolved.

- SAFETY SIGNIFICANCE:

Failure to perform testing of the generator at the nameplate rating of 3125 kVA .;

could result in a failure to detect degradation of the alternator, i.e. the ~

EDG's ability to carry accident loads. <

l t

't i

)

1 l

9x n ,

P

[

APPENDIX B ACRONYMS AND ABBREVIATIONS ,

A Amperes .

ACR Adverse Condition Report AECL Atomic Energy of Canada Limited A-H Amp-Hours j

AFW Aux 111ary Feed Water ANSI American National Standards Institute AOP Abnormal Operating Procedure ASTM American Society of Testing and Materials

-C Celsius -

CP&L Carolina Power & Light Co.

CVT Constant Voltage Transformer CWD Control Wiring Diagrams DBD Design Basis Document

. DCP Design Change Packages DBR Design Base Reconstitution DFOST Diesel Fuel'011 Storage Tank DR Deviation Reports DS Dedicated Shutdown EBASCO Licensee architect-Engineer EDG Emergency Diesel Generator EDS Electrical Distribution System  ;

, EDSFI Electrical Distribution System Functional Inspection EPM Electrical Planning and Management J ESF Engineered Safeguard Feature E&TS Engineering and Technical Support i EWR Engineering Work Request ff fouling factor ,

-j F0 Fuel Oil FR Field Report ]

GDC General Design Criteria i HBR-I H.B. Robinson coal-fired plant '

Hp Brake Horsepower HVAC j Heating Ventilation and Air Conditioning IEEE Institute of Electrical & Electronics Engineers l ISI Inservice Inspection  ;

JW -i Jacket Water KW kilowatts kV kilovolts kVA Kilovolt-amperes LERs Licensee Event Reports LO Lubrication Oil ,

i LOCA Loss of Coolant Accident LOOP- Loss of Off-site Power MCC Motor Control Center i MOV Motor Operated Valve i MVA Mega Volt-amperes

g_.___,.._

r o

?

Appendix B 2 MW Megawatts NED Nuclear Engineering Department NEMA National Electrical Manufacturers Association NSR Non Safety Related OST Operating Surveillance Test PIR Plant Improvement Request P&ID Process and Instrumentation Drawings POER Plant Operating Experience Report -

PM Preventive Maintenance PDM Plant Operating Manual '

PRA Probabilistic Risk Assessment PT Periodic Test.

RHR Residual Heat Removal RNP Robinson Nuclear Plant SAT Start-up Transformer SCR Significant Condition Report SW Service Water SP Special Procedure SQUG Seismic Qualification Users Group T&D Transmission & Distribution TS Technical Specification UFSAR Updated Final Safety Analysis Report VAC Volts Alternating Current VDC Volts Direct Current e

i b

-c -- i, - - ._

APPENDIX C

. PERSONS CONTACTED Licensee Employees

• G. Attarian Electrical, Unit Manager, NED

• T. Bowman: Engineer,.NED

• R. Chambers Plant General Manager,. Robinson Nuclear Project

• G. Chappell Civil-Mechanical Supervisor, NED

• C.'Coffman Engineer Technical-Support

• D. Dykesterhouse Engineer, NED

• S.' Farmer Manager, Engineering Programs, Technica1 ' Support

• W. Flanacan Manager, Operations

• S. Gupta Engineer, NED M. Hammack, Senior Engineer, NED

• B. Hynds Engineer, NED

• P. Jenny Compliance Engineer, Regulatory Compliance

, G. Kirven, System Engineer, Technical Support F. Legette Senior Control Operator, Operations

• A. Lucas Manager, NED .

• L. Lynch Supervisor Quality Control

• M. Macon Engineer, NED

• A. McCauley J. McDaniel, Manager, Electrical Maintenance, Technical Support Senior Engineer, NED

• R. Parsons Manager, Engineering Support, NED

~. *M. Payne Electrician First Class, Maintenance J. Pierce, Senior Engineer, NED

• R. Smith Manager, Maintenance

• K. Strouzas Engineer, NED

• G. Vaughn

• J. Windham Vice President Nuclear Services Electrician First Class, Maintenance

• Attended exit meeting

s _.

9 .

6 4

CP&L Carolina Power & Light Company ROBINSON NUCLEAR PROJECT DEPARTMENT

. POST OFFICE BOX 790 HARTSVILLE, SOUTH CAROLINA 29550 Robinson File No.: 13510E Serial: RNPD/92-0551 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 H. B. ROBINSON STEAM ELECTRIC PIANT UNIT NO. 2 DOCKET NO. 50-261 LICENSE NO. DPR-23 H. B. ROBINSON POSITION REGARDING CABLE SUBMERGENCE Gentlemen:

As requested in a telephone conversation between Mr. J. D. Kloosterman of my staff and Mr. M. B. Shymlock of your staff, Carolina Power and Light Company (CP&L) hereby provides a position on the submergence qualification of the Samuel Moore, BIV, and Continental cables installed at H. B. Robinson, Unit No. 2 (HER-2). As shown in the enclosure, it is CP&L's position that these

. cables have been and are still qualified for submergence at HBR-2.

Should you have any questions regarding this matter, please contact Mr. C. T. Baucom at (803) 383-1253.

Very truly yours

. H. Chambers General Manager Robinson Nuclear Project Department RDC/kac Enclosure cc: Mr. S. D. Ebneter Mr. L. W. Carner l

, Mr. M. B. Shymiock, -

l i

l L l

-'- i

- g' j$h ~

w .. .-.. - - - - - -

3= _ . - -

.g .

Enclecura to Stric1: RNPD/92-0551 y Page 1 of 6 CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PIANT, UNIT. NO. 2 POSITION ON SUBMERCENCE OUALIFICATION OF CABLE

REFERENCES:

1. " Samuel Moore Cable," HER-2 EQDP 11.1, Rev. 1 (March 8, 1991).

2. " Boston Insulated Wire (BIU) Cable," HBR-2 EQDP 36.0, Rev. 1 (June 14, 1991).

3. " Continental Cable," HER-2 EQDP 10.0, Rev. 1 (March 8, 1991)

4. " Submergence and High Temperature Steam Testing of Class lE Electrical Cables," NUREG/CR-5655 (May, 1991).

5. " Test Report for EGS GRAYBOOT Connectors, Models CB-1, GB-2 and GB-3," Report No. EGS-TR-880707-04 (November 26, 1990).

6. " Qualification of Okozel Insulation for Nuclear Plant Service," Okonite Report No. NQRN-4A, Rev. 1 .

(March 31, 1987).

7. " Performance Test of Raychem WCSF-N Splices on EPR /

Hypalon Wire," Raychem Report No. EDR-50ll, Rev. 1 (February 25, 1980).

8. " Qualification of Firewall SR Class 1E Electrical 1 Cables," Rockbestos.

The following is a discussion of H.B. Robinson Nuclear Power Plant's (HBR-2) position on the submergence qualification of the Samuel Moore, BIV.and Continental cables installed at HBR-2. It is CP&L's position that these ,

cables have been and still are qualified for submergence at HBR-2. In light of -

the recent release of preliminary information regarding cable submergence for Sandia (Reference 4 above), the following observations are provided.

1. Aeolicability of Sandia Tests The Sandia test was conducted to generic test profiles with a conservative generic activation energy being used. The cables provided by Samuel Moore for the Sandia test were aged for six months at 95'C prior to the design basis event simulation. Since the materials of construction for the subject cables are assumed to follow an Arrhenius relationship, the reqirements u at one time and temperature can be j transferred to another set of time-temperature coordinates using the  ;

following equation:

I l

a-. . . . - . . - - - - _ . .. . l

%-#m -- 1

-q - .

l Encleruro to Ssrial: -RNPD/92-0551  !

g- Page 2 of 6 l t2 - t, exp { (E,/kg)(1/T,-1/T2 )]

where, t2 - equivalent time at T2 t, - time at temperature T, E. - activation energy (eV)- (1.31 eV for Samuel Moore cable used at HBR-2, Ref. 2) kg - Boltzmann's Constant (8.617x10-5 eV/K)

T, - accident temperature (K)

T2 - reference temperature (K)

Utilizing the aging information in the above equation along with the activation energy of 1.31 eV (Ref. 1) and the HBR-2 service temperature of 120*F (49 *C), this yields a qualified life of 130 years. This exceeds.the HBR 2 requirements by a factor of 3, indicating that the test cables were significantly more degraded than the HBR-2 cables will be at the end of their plant life.

Similarly, the same Arrhenius equation used to determine qualified life  ;

can be used to perform a time-temperature equivalency to show that the Sandia accident simulation exceeded the HBR-2 plant accident profile by significant margin. Provided below are the two profiles to be compared:

1 SANDIA PROFILE HBR-2 PROFILE T2 - 152 *F (Ref. Temp.) T2 - 152 *F

- t - 4 hrs t2 - 2.78 hrs T - 327 *F T - 265 *F ,

t, - 2 hrs t - 21.22 hrs Ta

- 300 *F T3

- 219 *F t.

- 220 hrs t.

- 696 hrs T. - 246 *F T. - 152 *F E. - J1.31 eV Solving the Arrhenius equation provides an equivalency of 7606 days for-the Sandia LOCA profile as compared to 227 days for the HBR-2 profile.

It can again be seen that the Sandia test subjected the Samuel Moore cables to a si nificantly 5 more' severe environment than required for HBR-2. It should also be noted that after being subjected to very severe testing conditions, the Samuel Moore cables passed the post-LOCA dielectric withstand tests and in fact remained functional until sometime during the submergence test.

1 The testing performed by Sandia on the BIW cable supports the use of i this cable in submerged applications. Using the same Arrhenius l equation, service temperature, and aging parameters described above along with the appropriate activation energy (1.15 eV, Ref. 2) for the BIW cable yields a qualified life of 66 years.

.. - - . ....n.. ._._

y .. . . . .. ..

_ _ - _ _ - _ = _

E bx r

L . Enc 1ccura to Ssricl: RNFD/92-0551 Lg Page 3 of 6 The LOCA degradation equivalency was also performed to HER-2 parameters using the same method as shown for the Samuel Moore cable with the only change being in the ectivation energy, 1.15 eV for BIW cable (Ref. 2).

Solving the Arrhenius equation provides an equivalency of 2943 days for

! the Sandia LOCA profile as compared to 125 days for the HBR-2 profile.

This demonstrates that the Sandia test subjected the BIW cable to a more severe environment than required for HER-2. The BIW cable successfully passed the Sandia submergence test (see Ref. 4, Table 7., Multiconductors 10 & 11). These conductors also passed the post-LDCA dielectric withstand test (see Ref. 4. Table 8).

The Sandia test did not contain a sample of the Continental wire. CP&L believes that the Firewall SR cable tested by Sandia closely resembles our Continental cable as shown below (Reference 3).

Continental Cable 2/C, 16 AWG, 7 strand coated copper conductor, .030" silicone rubber insulation, color coded glass braid over insulation, asbestos filler, cabled, mylar binder tape,1/C,16 AVG, 7 strand I uncoated copper drain wire, copper backed mylar tape, metal side in contact with drain wire, glass braid, .045" silicone rubber

, jacket. .

Rockbestos had itt Firewall SR cable tested (Ref. 4, Supplier 10) which is similar in construction to the Continental cable, and the Firewall SR  ;

cable successfully completed the submergence test. Information provided in the Rockbestos catalog indicates that the Firewall SR cable is a multi-conductor shielded cable as described below:

Firewall SR Cable 2/C, 16 AWG 7 strand tin-coated copper conductor, silicone rubber insulation, glass braid, tape shield, tin coated copper drain L wire, binder tape, and Rockhide jacket.

l Aging of the Firewall SR cable (by Sandia) under HER-2 service conditions yields a qualified life of 716 years (using an activstion energy of 1.71 eV, Ref. 8). Also, solving the Arrhenius equation provides an equivalency of 1.03x10s days for the Sandia LOCA profile as i compared to 1.36x10 8 days for the HBR-2 profile. This demonstrates that l the Sandia test subjected the Firewall SR cable to a more severe  ;

I environment than required for HBR-2.

l As can be seen by the above discussions, the Sandia tested cables were l aged to environments significantly more severe than postulated for l HBR 2. i e*

Eoclocuro to Scrial: RNPD/92-0551 j Page 4 of 6

2. Cable Construction The second observation is that of the difference in cable construction of the Samuel Moore cable tested by Sandia versus that used at HBR-2.

Tested Samuel Moore Cables:

Supplier No. Description (Table 1 / Ref. 4) 4 Dekoron Polyset, XLP0 Insulation, CSPE Jacket, 12 AUG, 3/C and Drain 7 Dekoron Dekorad Type 1952 EPDM Insulation, Individual CSPE Jackets, Overall CSPE Jacket, 16 AVG 2/C TSP 600 V Samuel Moore Cable at HBR-2:

Descriolion 2/C, 16 AWG, 7 strand tinned copper conductors, 20/10 mil EPDM/Hypalon insulator - conductor jacket, aluminum -

mylar shield, 45 mil Hypalon jacket The construction of the BIV cable is identical between the Sandia test sample and that used at HER-2. The similarity of the Continental cable has been discussed previously.

1

3. Cable Testinz The final observation concerning the Sandia cable test is the severity of the dielectric withstand test. The Sandia test subjected'the cables to a dielectric withstand test prior to the submergence test. Sandia has indicated that the IEEE-383 dielectric withstand tests are very severe and can induce failure in otherwise functional cables.

4. Vendor Soonsored Test Programs l HBR-2 qualification of the subject cables is based on vendor sponsored ,

test programs as identified in References 1, 2, and 3. The following l demonstrates that the subject cables, fully functional at the end of )

their respective test programs, exceed the HBR-2 requirements by a wide f margin.

.,www.w #e== wee. mopew. - w....

%a .

L o.

L Enclecura to S::rinl: RNPD/92-0551

• Pego 5 cf 6 .

9.

The' tests referenced in HBR-2 EQDP 11.1 (Ref. 1) over-tested the Samuel.

Moore cables when compared to HER-2 requirements as follows: ,

Aging: =107 years @ 49 *C IDCA: 12,364 days to HBR-2's 227 days (degradation L

equivalency @ 152 *F) ,

i IR Data: 2.9x10' ohms @ 300 *F and 5.8x107ohms @ 200 *F

/ 30 day test (Isomedix)> 1x10e ohms @ 223 *F & ,

10 psi / 100 day test (NTS)  ;

Submergence: Section 3.4.7 of Ref. 1 describes submergence testing performed by Samuel Moore on cables for 52 weeks in 90 *C water. Measurements of ,

I Specific Inductive Capacity & Power. Factor were taken to demonstrate the electric stability of the cable while submerged. l The tests referenced in EQDP 36.0 over-tested the BIU cable when compared to HER-2 requirements as follows:  !

Aging: =43 yrs. @ 90 *C (conservative ref. temp.) & '

4718 yrs. @ 49 *C  ;

i IDCA: 5,803 days to HBR-2's 125 days (degradation equivalency @ 152 *F)

IR Data: 2.9x10s ohms @ 300 *F and 8.4x107 ohms @ 200 *F/ i 45 day test (BIV)ICEA Test for 52 weeks @ 75*C l yielded 7.9x10' ohms (lowest value for 52 weeks)

The test referenced in EQDP 10.0 was specific test for HER-2 and subjected the continental cable to a full environmental qualification test as described:

Aging: =41.8 yrs. @ 49 *C LOCA: 24,552 days to HBR-2's 7,359 days (degradation equivalency @ 152 *F)

IR Data: 2.8x107 ohms @ 282 *F The manufacturer provided submergence test data for the subject cable. I Accelerated Water Absorption testing was performed for 14 days in 75 *C water & for 7 days in 75 *C water. These tests showed that silicone rubber is inherently water resistant and that change in dielectric strength of the cable remained essentially unchanged.

i l

. , - , . , - - :1 - - . . . - . - - - . . . - - . . - - . . . -

H _

.l i

e .. -.

inciccurs to Sarici: RNPD/92-0551  !

g) Page 6 of 6 Additionally, Reference 5 documents testing performed on twenty-five i wire samplos to a full sequential qualification test program including a-  ;

30-day submergence test. The wires were from various manufacturers and i included XLPE, EPDM and silicone insulation materials. Each sample, approximately 25 - 30" in length, was contained within the test chamber. .

Five of the samples exhibited problems with wire insulation' splits and {

cracks which was a result of handling damage during the ,

thermal / radiation aging. The other 20 samples exhibited equivalent irs  :

(insulation resistance) throughout the LOCA (with peak temperatures and l pressures of 354 'F / 77 psig) and the 200 *F submergence test.  !

)

A CP&L review was also performed of other qualification tests focusing on IR data obtained during LOCA and post-LOCA submergence of various cable samples (Refs. 6 and 7). This review indicated that IR values .

during LDCA and post-LDCA submergence, after appropriate correction to a i common reference temperature, are equivalent (i.e., within the same  ;

order of magnitude). This principal also applies to the Sandia test j results. Even though the data on performance of the Sandia test samples during the LOCA simulation has not yet been released, the IR values  ;

exhibited during the submergence test are within the expected range i based on comparison to similar EQ LOCA steam tests. Also, exposure to  ;

caustic chemical spray solutions is not necessarily more degrading than  !

. . exposure to distilled water (which results in worse swelling).

NUREG/CR 4301 exposed polymer materials to water & IEEE 323-1974 chemical spray solutions, concluding that swelling in boiling water was approximately twice that in chemical spray solutions.

Therefore, based on the analysis presented above, CP&L maintains that .

the HBR-2 installed cables discussed above are qualified for submergence  !

applications at HBR-2. CP&L will continue to monitor and review the v

'Sandla test program for further information on submergence qualificatior-k i

a 1

l w- - . - . - . - - - , . __ _ _