ML18018B476

Forwards Responses to Draft SER Open Items Re Instrument Air Sys,Nsss Startup Testing & Load Testing of Diesel Generators

ML18018B476
Person / Time
Site:	Harris
Issue date:	11/15/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
LAP-83-532, NUDOCS 8311220523
Download: ML18018B476 (24)

Text

I REGULATORY INFORMATION DISTRIBUTION SYST (RIDS)

ACCESSION ABR:8311220523 D

~ DATE: 83/11/15 NOTARIZED, NO FACIL:SO-400 Shearon Harris Nuclear Power Planti Unit 1< Carolina 50-40$. Shearon Harris Nuclear power Plant~ Unit 2~ Carolina AUTH ~ NAME AUTHOR AFFILIATION MCDUFFIEpMeA ~

Carolina Power 8, Light Co, REC IP ~ NAMC RECIPIENT AFFILIATION DENTONgH ~ Rr Office of Nuclear Reactor Reg'ulationi Director A

DOCKET 05000400 05000401

SUBJECT:

Forwards responses to draft SER open items re instrument air sys~NSSS star,tup testing 8 load testing of diesel generators, DISTRIBUTION CODE!

0001S COPIES RECEIVED:

TITLE: Licensing Submittal:

PSAR/FSAR Amdts L Related Correspondence'OTES:

RECIPIENT IO CODE/NAME NRR/OL/AOL NRR LB3 LA COPIES LTTR ENCL 1

0 0

RECIPIENT COPIES IO CODE/NAME L.TTR ENCL NRR L83 BC 1

0 BUCKLEY'S 01 1 1

INTERNAL: ELO/HDS1 IE/DEPER/EPB 36 IE/DEQA/QAB 21 NRR/DE/CEB 1$

NRR/DE/EQB 13 NRR/DE/MEB 18 NRA/DE/SAB 24 NRR/DHFS/HFEB40 NRR/OHFS/PSRB NRR/DSI/AEB 26 NRR/DS I/CPB 10 NRR/OS I/ICSB 16 NRA/OS I/P8 8 19" NRA/DSI/RSB 23 RGN2 EXTERNAL: ACRS 41 DM8/DSS (AMDTS)

LPbR 03 NSIC 05 0

3 1

1 2

1

,1 1

1 1

1 1

1 3

6' 1

1 1

1 1

IE F ILE IE/DEPER/IRB 35 NRR/DE/AEAB NRR/DF/EHEB NRR/OE/GB 28 NRR/DE/MTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NRR/DL/SSPB NRR/DS I/ASB NRR/OS I/CSB 09 NRR/DSI/METB 12 NRR/D RAB 22 FI'LE 04 AMI/MIB BNL(AMDTS ONLY)

FEMA REP DIV 39 NRC PDR 02 NTIS 1

1 1

1 0

1 1

2 2

1 1

1 1

1 1

1 0

1 1

1 1

1 1

1 1

1 1

0 1

1 1

1 1

1i TOTAL NUMBER OF COPIES REQUIRED; LTTR 53 ENCL ~

~

v 'I' V

r I V

I I

(

d

~

II

(

~

~

1

('III I

I I(r rd, I I I 11

(

Hrr bHV

')H kd

~I)

'Ir" dk

'I I

~

f( << f II ri

~<<w

<<(r<<

Htf g)

H X3 l

1 I

HI>>

V HX r<<

I I<< II'I ll i

~ 'I,

<<s 1~ i(V I

Hi r

<<Mr

CMtt, Carolina Power & Light Company ROY Is tg8g SERIAL:

LAP-83-532 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS.

1 AND 2 DOCKET NOS ~ 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT RESPONSES

Dear Mr. Denton:

Carolina Power

& Light Company (CP&L) hereby transmits one original and forty copies of responses to Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report Open Items.

The response numbers are listed on the cover page of the attachment along with the corresponding review branch and reviewer for each response.

We willbe providing responses to other Open Items in the Draft Safety Evaluation Report shortly.

Yours very truly, CT/cfr (8499NLU)

Enclosures M. A. McDuffie Senior Vice-President Nuclear Generation CC:

Mr. B. C. Buckley (NRC)

Mr. O.

Chopra (NRC)

Mr. H. Li (NRC)

Mr. W. Long (NRC)

Mr. E. Marinos (NRC)

Mr. G. F. Maxwell (NRC-SHNPP)

Mr. J.

P. O'Reilly (NRC-RII)

Mr. Travis Payne (KUDZU)

Mr. Daniel F. Read (CHANGE/ELP)

Mr. R.

P. Gruber (NCUC)

Chapel Hill Public Library Wake County Public Library Mr. Wells Eddleman Dr. Phyllis Lotchin Mr. John D. Runkle Dr. Richard D. Wilson Mr. G. 0. Bright (ASLB)

Dr. J.

H. Carpenter (ASLB)

Mr. J. L. Kelley (ASLB) 83ii220523 83iii5 PDR ADOCK 05000400 E

PDR 411 Fayetteville Street o P. O. Box 1551 o Raleigh, N. C. 27602

~

~

~

~

AA )

I

~

(1g m

W la Ja y4 r

J 1

i'

LIST OF OPEN ITEMS REVIEW BRANCH AND REVIEWER Procedures 6 Systems Review Branch/W. Long Open Item 202 Reactor Systems Branch/E. Marinos Open Item 235 Power Systems Branch/0.

Chopra Open Item 389

~

e

Shearon Harris Nuclear Power Plant Safety Review Questions:

640.21, 640.7, 640.36, 640.9, and Open Item 202 Revised Res onse Safety review questions 640.21, 640.7, 640.36, 640.9, and 640.5 (OI-202) were submitted to CP&L on June 22, 1982.

The initial responses were submitted to the NRC via Amendment 4 to the FSAR.

Based on additional discussion with the NRC staff, CP&L understands that additional clarification to these responses is required.

This clarification is provided below.

Response

1.

Question 640.21 The response to "this item was an amen'dment to 'FSAR Section 14.2.12.4.25.

This revision included an error which will be corrected in a future amend-ment; the revision to -be made is shown on the attached mark-up of FSAR page 14.2.12-29.

2.

Question 640.7 The description of the test summary for the instrument air system is found in FSAR Section 14.2.12.1.79.

This section will be amended to indicate that the sudden loss of pressure test will be conducted by isolating the instrument air system and bleeding off the air in the segment.

The FSAR will be revised in a future amendment as shown on the attached FSAR page

14. 2. 12-7.

With regard to the Regulatory Guide 1.68.3 (Rev. 0), Position C.10, the instrument air sys'em does not contain any single large loads which would cause a significant perturbation on the normal instrument air pressure.

Therefore, a test to verify the instrument air system's response to the conditions postulated in C.10 will not be conducted.

A revision to FSAR Section 1.8, Reg.

Guide 1.68.3, will be made in a future amendment to address this issue.

The response to this item was an amendment to FSAR Section

14. 2. 12. 2, 17.

This section will be revised in a future amendment to clearly state that NSSS startup testing will be performed using step-load increases and decreases.

The FSAR will be amended as shown on the attached FSAR page

14. 2. 12-89.

The response to this item was an amendment to FSAR Section

14. 2. 12. 1, 16.

This section of the FSAR will be revised in a future amendment to clearly state that load testing of each diesel generator will be in accordance with positions c.2.a (3) and (5) of Regulatory Guide 1.108, Revision 1.

The FSAR will be amended as shown on the attached FSAR page 14.2.12-19.

5.

Open Item 202 The response to this open item was submitted to the NRC in a letter dated July 22, 1983.

The response addressed the participation of licensed operators in the speciaL low-power testing program.

The testing program to be used is the revised test sequence proposed by Westinghouse in a letter dated July 8, 1981; this sequence was approved for use at facili-ties employing a Westinghouse NSSS in a letter dated October 23, 1981 from the NRC to Westinghouse.

FSAR Sections

14. 2. 12. 2. 26 and

14. 2. 12. 2.81 will be amended in a future amendment to include test methods for the special low-power testing pro-gram.

The FSAR will be revised as shown on the attached pages.

The revisions address items 1,

2, 3, and 5 in the test series.

Item 4 is currently addressed by FSAR Section 14.2. 12.2.21.

SHNPP FSAR 2)

The response times of the pump and valves are in accordance with FSAR Section 6.3.

3)

The residual heat removal pumps can supply adequate suction head to the Charging/Safety Injection pumps.

')

Critical valves are Wable to open against the maximum expected differential pressure conditions as specified by the NSSS supplier.

5)

Safety Injection components actuate to the state specified in FSAR Section 6 ~ 3.2.1 on receipt of a safety injection signal.

6)

The measured charging/safety injection pump NPSH meets or exceeds that required by the vendor technical manual for the CSIP.

14.2.12.1.26 High-Head Safety Injection System Check Valve Test Summary a)

Test Objective 1)

To demonstrate that emergency core eood.ing water can be delivered into the reactor coolant system at approximately no-load operating temperature and pressure.

b)

Prerequisites 1)

Hot Functional Testing in progress with the RCS at approximately no-load pressure and temperature.

2)

The water level in the pressurizer is as low as practical.

c) 3)

The applicable 'general prerequisites are met.

Test Method 1)

Injection shall be verified by flow noise and/or increase in pressurizer level for each cold leg inject'ion path.

d)

Acceptance Criteria 1).

The Safety Injection System injects emergency core cooling water into the reactor coolant system at approximately no-load operating temperature and pressure.

14,2,12 ' '7 Safety Injection Accumulator Test Summary a)

Tes t Objective

• 1)

To verify the discharge characteristics of each Safety Injection Accumulator.

I 2)

To demonstrate that the SI Accumulator discharge isolation valves

~ will open under the maximum differential pressure conditions.

14.2.12-29 Amendment No.

4

SHNPP FSAR Acceptance Criteria I)

The automatic temperature control of the sampled fluids is regulated per Ebasco specification 2)

The Secondary Sample Cabinets instnmentation and alarms function per FSAR Section 9' '.5 and Ebasco Specification.

3)

Sample points have been verified per FSAR Table 9.3.2.1.

4)

Operation of the "Hotwell Sample" pumps are in accordance with system design wiring diagrams and Ebasco Specification.

14 ~ Zo 12. I ~ 79 Loss of Instrument Air Test Summary a)

Test Obgectives b)

I)

To demonstrate that a reduction and loss of instrunent air

,.pressure causes.,fail-.,safe,.operation of pneumaticalgy-operated valves and rjampers bogh safety and nonsafety relatedgcccl4 ~+~

bv.rig:~g) amdt<'sr) 4u-iSoy ~M fur( k 4<:y 6 i~4:~q

~

Prerequisites O

1)

The general prerequisites are met.

2)

Specific prerequisites will be delineated in the system preoperational test procedure.

c)

Test Method I)

Qhere safe to personnel and equipaent, a slow reduction in pressure and a loss of pressure test will be performed.

Testing will be done in small segments/individually and response noted for both safety and nonsa et gelated valves and dampyrs T~ '1~45 o

p~~~~~

0+1 ~ i( 4~~

eM bp is la%;~ cqw~p<;~;guZ Qc~~

,Acceptance Criteria V~r'g ~ ~)c Pram fhe isola'j

~~~

I)

Proper fail-safe operation of valves and dampers subject to a reduction and loss of instrument air is verified.

14.2 12.1.80 Containment Building Hot Penetration Testing a)

Test Ob)ective I)

To demonstrate that containment concrete temperatures in the vicinity of hot containment penetrations do not exceed 200F.

b) prerequisites

'I)

.he general prerequisites are met.

2)

For hot containment penetrations, the designer has provided the anticipated locations of maximum concrete temperatures.

14.2.12-77 Amendment Ho.

7

SHNPP FSAR 4)

,Perform a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> load reliability test of each diesel generator unit~ a-adorj~wae.

w i 'fk R4 1-1o 8 CQm. I) C.Z,.4 (~) 4W d) 5)

Perform a Combined Mode Diesel Generator Unit test to demonstrate independence of the two units.

Acceptance Criteria 1)

E h di 1 generator starts (automatic start or Local/reraote ac ese manual start) and accelerates to rated speed and voltage w

n seconds after receiving a signal to start.

2)

E h di 1 generator shall start and accelerate emergency bus ac ese loads to rated speed in the required sequence (FSAR Table 8.

~ " )

without exceeding five percent speed

drop, 20 percent voltage drop at the load terminals, and 'recover 'to within 10 percent of nominal voltage and two percent of nominal frequency withip 40 percent of each load-sequence time interval.

3)

The diesel generator speed shall not exceed 107.5 percent of nominal (450 rpm) upon disconnecting of the single largest load and shall not trip on overspeed (110 percent of nominal) upon disconnecting of all load.

4)

The diesel generator and auxiliaries operate to mainta intain IOOX load for a 22 hour2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> period and IIOX load for an additional two hours.

5)

Each diesel generator satisfactorily performs during thirty-five consecutive starts per Regulatory Guide 1.108, Section c.2.e.

6)

Both diesel generators start and operate satisfactorily upon receiving simultaneous start signals.

14.2 ~ 1.2.1 ~ 17 Xncore Nuclear instrumentation Test Summary a)

Test'Obgective To demonstrate that the incore monitoring system functions to remotely position the incore neutron detectors for the purpose of flux mapping.

b)

Prerequisites I)

Cleaning operations and complete.

tubing connection verifications are 2)

The drive unit has been loaded with a dummy detector.

3)

The general precautions are met.

14 ' '2-19

.Amendment No,

S'

SHNPP FSAR 2)

Using data results, and assumed calorimeter information, calculate the average steam generator moisture carryover.

d)

Acceptance Criteria Verify with a radioactive tracer in5ection method that steam generator moisture carryover is no more than 0.25X at the per FSAR 'fable 10.2.1-1 steam generator output.

14.2 '2.2.17 Load Swing Test Summary a)

Test Obgective b) 1)

To verify proper nuclear plant and secondary plant transient

response, including automatic control system performance when load changes are introduced at the turbine generator.

Prerequ'isites I

1)

Plant conditions are established as required by the test instruction.

I 2)

The general prerequisites are met.

c)

Test Method Note:

Step load changes~will be initiated from steady state conditions at approximately 30, 75, and 100 percent power in accordance with Vestinghouse NSSS Start"up Manual Section SU-3.4.7.

l~ c~

1)

Manually initiate a in the turbine generator output as rapidly as possible to achieve an approximate 10 percentg1.oad

~C 2)

Plant variables will be, recorded, along with values observed on normal plant instrumentation, during the load transient for those parameters required.

d)

Acceptance Criteria 1)

The following acceptance criteria, are to be used to determine successful test completion.

Failure to meet those criteria does not constitute a need for stopping the test

program, but correction of any deficiencies should be accomplished as required consistent with the current plant schedule.

(a) 1.1 Reactor and turbine do not trip.

(b) 1.2 Safety infection is not initiated.

(c) 1.3 Neither steam generator relief valves

@or safety valves lift.

14.2.12-89 Amendment No.

4

~ ~. i>.

~ a ~ ~ Cu tV ~. I C.

SHNPP FSAR 14 ~ 2 ~ 12 ~ 2 ~ 26 Natural Circulation Test Summary a)

Test Ob]ective 1)

To confirm that design heat removal capability exists under natural circulation conditions

~

2)

To verify that flow (without pumps) and temperature data are comparable to prototype designs for which equivalent tests have been successfully completed.

b) c)

3)

To obtain a data base for simulator training in natural circulation operation.

g vs dyncygrak p/g iAh&~f k~s cfpre~uri ~em J)ert +e, n<< i ~~)'~Q

~ '8' Prerequisites >~~nc f/oa yn maroin + za j'cu-+inc ~per>~.

P/~ ig~/e Q appreximak4y 8</~ ~psramekr~a% no-/M value.'~.

,1 i)

~

~I Wee reawoke ~land Pumps are rcrnnin'.

2) t r.

3)

The 9~nxC prere~deks are mes Test Method P

alp %c rcaic*>> cao/and purnpa.

1)

At 5X rate thermal powe the plant in nat circu tion ode recording ength of ti for the t to s t live, th r bution, and the abi raaintain co g

'eai~ aJe~k nct~g g/rgu/wPIivc f lorn yb ~~g~'/z '~.

2)

Perform s of-Offsite Pbwer/St Blackout Test with ant trip fro

-20X ated Thermal Po

~

Oper te plant estab shin stabl ondition in natural culation usi g batteri and emer ency di ls.

Record he lengt of time for the ant t stabilize, th ow distribution, e ability to maintain ing mode d)

Acceptance Criteria J 5'ce/ ~

P 1)

The heat removal capability, flow~ and temperature data, oac~ae are comparable to those of the prototype5 (North Ann for which equivalent tests have been completed.

2} 5 res,sor'izing.

F ~Rue Mrna'~ above l800 prig.

14 ~ 2 ~ 12 ~ 2 ~ 27 Main Steam and Feedwater Systems Test Summary a)

Test Ob]ective 1)

To verify the operating characteristics of the main steam and feedwater systems during power escalation.

b)

Prerequisites 1)

Power. level is established as necessary to meet the test requirements.

2)

The general prerequisites are met.

I B. Ob~~c ~g~r JjgkrjbRk IcH.~

~

~~~+4k 6c 54z4 nd ln~raA dtrK&gfvvrlr4454d k) rk rrreras~lAe~J a

Lt

r4 ~. r>.~.+c dAaApunp-~

I J

I I

1

I

/4s. ra;/Z

')-~~~~~ w ~u'w~

m~Cg ~

.... c5ttZ~ +@a~ -~~

g ~<

~8~

)

ACZd ~<~,

A)~

eke ~~

i)

+~i~~

y p

43%2/,.SUCAV,,Ag~eo 'uic&u d

~ CR8P7f ~~ p~~~

g~ Accanrr-aA~u

~) cue~ 2a. ~uird'~

Shearon Harris Nuclear Power Plant Draft SER Open Item 235 TMI A endix a

e TMI-53 Provide a response to NUREG 0737 item II.K.3e17, "Report on Outages of Emergency Core Cooling Systems Licensee Report and Proposed Technical Specification Changes."

~Res ense:

In the first five years following issuance of an operating license for SHNPP, Unit 1, CP&L will collect the following data on ECC systems and components:

1.

s2 ~

3e 4.

Outage dates and duration of outages Cause of each outage ECC systems or components involved in the outage If applicable, corrective actions required Test and maintenance outages willbe included in the data collection.

A report of the data collected will be submitted to the NRC within one year of the end of the five-year period.

The FSAR willbe revised in a future amendment to reflect this response.

{8506COM/ccc)

Shearon Harris Nuclear Power Plant Draft SER Open Item 389 Su lemental Information The NRC (0. Chopra, Power Systems Branch) has requested that CPGL provide additional information on short circuit protective devices for circuits that pass through containment penetrations.

RESPONSE

Figure 1 depicts the typical system arrangement for the Auxiliary System at Shearon Harris Nuclear Power Plant.

Under normal conditions if there was a fault to ground (see Figure

1) on the, secondary side of the control power transformer.located in the Hotor Control Center, the fuse located in the secondary side would blow and protect the electrical penetration.

Under the most unlikely event where the fuse fails to clear the fault the transformer impedance is great enough to limit the short circuit current to a level which the containment penetration can carry without damage.

The scenario requested to be considered by the reviewer is one in which there is a ground fault in containment on the control cable connected to the control power transformer, the fuse fails to interrupt and finally the control power transformer fails due to over-heating caused by it carrying a higher than normal current.

The way the NRC postulates this failure the 480V side of the transformer being directly connected to the secondary side of the transformer thus applying 480 volts to the control circuit.

The 6.9 kV 480V transformer which is the source of 480V power has its 480V secondary neutral connected to ground through a resistor.

This grounding resistor is sized to limit ground fault current to 15 amps.

Therefore, in the above scenario where 480 volts is directly applied to the, control circuit, the short circuit current would be limi'ted.to 1'5.amps.which is within the capability of the penetration.

See Figure (2).

Therefore, we can conclude that damage to the penetration does not occur.

(8509N LUccc)

V

\\

e

~ ~ c,eg e.elle J',e e

e id'er ~~~ pe eeet>>

~'JAAUf~4rvAIJL44~i~>>&ALL&:?i e'NRwAl e&k)4~e P)e'e.)'oSvtaeeeee All,tl(8

~ ~ ~

EBASCO SERVICES 1NCORPORATEO BY c

CHKOe BY CLIENT DATE DATE o ~~A Pa~

OFS HO.

SHEET OF DEPT.

HO.

PROJECT SUBJECT Fcs u 0

C cW 4g Kv Qug i

~ ~

e e

'I8oM Gu&

e

)

qCteeWI Paeeor 7e~5 &eeeee e

e t80 lQog Pen+

~ '

~

~

~

~

~

1

=T

e

~

V I

8 n

ASCO SER lt'ICES !INCORPORATED CNlCD. BY DATE ET SATE OFEI NO.

SHEET OF DEPT.

NO.

CLIENT OWMA I 0~

PROJECT SUBJECT gg KV Qgc

I A

It'