ML18018A509

Forwards Response to 830208 Generic Ltr 83-10C Re TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pump. Util Will Participate in Westinghouse Owners Group Program to Address Issue

ML18018A509
Person / Time
Site:	Harris
Issue date:	04/22/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
TASK-2.K.3.05, TASK-TM GL-83-10C, GL-83-10D, LAP-83-117, NUDOCS 8304270311
Download: ML18018A509 (14)

Text

REGULATO INFORMATION DISTRIBUTION TEM (RIDS)

ACCE'SSION NBR;8304270311 DOC,DATE: 83/04/22 NOTARIZED:

YES FACIL:50-400 Shearon Harris Nuclear Power Plant~ Unit 1~ Carolina 50-401 Shearon Harris Nuclear Power Plant~ Unit 2~ Carolina AUTH BYNAME AUTHOR AFFILIATION MCOUFFIEqM>A, Carolina Power 8 Light Co ~

RECIP ~ NAME RECIPIENT AFFILIATION EISENHUT~DiG, Division of Licensing S UBJECT:

For war ds response to 830208 Generic Ltr 83 10C r e 'TMI Action Item II.K'.5i "Automatic Trip of Reactor Coolant-Pump," Util will participate in Nestinghouse Owners Group Program to address

issue, DISTRIBUTION CODE B001S iCOPIES RECEIVED ~ LTR ENCL g, SIZEe.

TITLE: Licensing Submittal:

PSAR/FSAR Amdts 8, Related Correspondence NOTES:

DOCKET 05000400 05000401 REC IP IENT ID CODE/NAME NRR/DL/ADL NRR LB3 LA INTERNAL: ELD/HDS1 IE/DEPER/EPB 36 IE/DEQA/QAB 21 NRR/DE/CEB 11 NRR/DE/EQB 13 NRR/DE/HGEB 30 NRR/DE/MTEB 17 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NRR/DSI/AEB 26 NRR/DSI/CPB 10 NRR/OSI/ICSS 16 NRR/DS I/PSS 19 NRR/DSI/RSB 23 RGN2 EXTERNAL: ACRS 41 DMB/DSS (AMDTS)

LPDR 03 NSIC 05 RECOPIES LTTR ENCL 1

0 0

1 0

3 1

1 1

2 2

1 1

1 1

2 2

1 1

1 1

1 1

1 1

1 1

1 3

6 1

1 1

1 1

RECIPIENT ID CODE/NAME NRR LB3 SC KADAMSIt P 01 IE FILE IE/DEPER/IRB 35 NRR/DE/AEAB NRR/DE/EHEB NRR/DE/GB 28 NRR/DE/MEB 18 NRR/DE/SAB 24 NRR/DHFS/HFEB40 NRR/DL/SSPB NRR/DS I/ASB NRR/OSI/CSB 09 NRR/DSI/METB 12 NRR/~

AS 22 04 IB BNL(AMDTS ONLY)

FEMA REP DIV 39 NRC PDR 02 NTIS COPIES LTTR ENCL 1

0 1

1 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

1 1

TOTAL NUMBER OF COPIES REQUIRED:

LTTR 54 ENCL 47

~ t I

<<\\

<<i It ii i<<,

il f I'I I

III I

'l'Iat' a

~

"~

I/(~

I ilfI )

"hf l

'1 it) PI <<'I 'I<<("<<I

<< f<<l I

I,ii P

li

'QI <<IPttti II<<AI jl II yf t/r; Af a

". )sl<<

O

> (0 Pt tq I, ffaa, Iv "yg iq ~

Ii I I/R '[)<

i'\\

r, f

isa a

il I'<<I

g If il R'

IN I

I)+II I]

g 4

I 1~+

I <<

'Pt,t

'I I'4,',

/,I PIXt

'Ptl,'

'Pl, 7f

, rll "giifl II'll

)

R I

I R

Itl Kg 1

R P iil X N l

't

~

'll i Pl R

r, If I/I <<R J~ti '.I ~

a R

CO)QE, Caroanaggca[ ftgf Company SERIAL:

LAP-83-117 Mr. Darrell G. Eisenhut, Director Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOSa 1

AND 2 DOCKET NOS ~ 50-400 AND 50-401 RESOLUTION OF TMI ACTION ITEM II.K.3.5 "AUTOMATIC TRIP OF REACTOR COOLANT PlBPS" (GENERIC LETTER NO.83-10c)

Dear Mr. Eisenhut:

Please find attached Carolina Power

& Light Company's (CP&L) response to your letter of February 8, 1983 concerning TMI Action Item II.K.3.5, "Automatic Trip of Reactor Coolant Pumps."

Carolina Power Light Company is a member of the Westinghouse Owners'roup (WOG) and as such will be participating in the MOG program to address this issue.

That participation is reflected in the attached response.

As requested by your

letter, our response is submitted under oath.

Tf you should have any questions on the attached

response, please do not hesitate to contact our staff.

Yours very truly, M. A. McDuffi.e Senior Vice President Engineering

& Construction JJS/mf (6684JJS)

Attachment Mr. N. Prasad Kadambi (NRC)

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

Mr. J.

P. O'Reilly (NRC-RII)

Mr. Travis Payne (KUDZU)

Mr. Daniel F.

Read (CHANGE/ELP)

Chapel Hill Public Library Make County Public Library Ny nome%asian expires:t Sl/ /rD CC:

Mr. Wells Eddleman Dr. Phyllis Lotchin Ms. Patricia T. Newman Mr. John D. Runkle Dr. Richard D. Wilson Mr. G. 0. Bright (ASLB)

Dr. J.

H. Carpenter (ASLB)

Mr. J. L. Kelley (ASLB) yr a

M. A. McDuffie, having been first duly sworn, did depose and say that the

information contained herein is true and correct to his own personal knowledge or based upon information and belief.

r F

<<<<~yetteville Street o P. O. Box 1551 o Raleigh, N. C. 27602 83042703ii 830422

,'PDR ADOCK 05000400

. PDR

~

~

C

CAROLINA POWER

& LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT PLAN FOR RESOLUTION OF TMI ACTION ITEM II+K.3.5 "AUTOMATIC TRIP OF REACTOR COOLANT PUMPS" INTRODUCTION The criteria for resolution of TMX Acti'on Plan Item II.K.3.5, "Automatic Trip of Reactor Coolant Pumps" were stated in letters from Mr. Darrell G. Eisenhut of the Nuclear Regulatory Commission to all Applicants and Licensees with Westinghouse designed Nuclear Steam Supply Systems (83-10 c and d) dated February 8, 1983.

The following represents the plan for demonstrating compliance with those criteria.

In order to avoid confusion, the overall philosophy and plan will first be stated.

Then, each section of the attachment to NRC letters 83-10 c and d will be addressed as to how the overall plan responds,to each NRC criteria.

OVERALL PLAN In the four years that have passed since the event at Three Mile

Island, Westinghouse and the Westinghouse Owners'roup have held steadfastly to several positions relative to post accident reactor coolant pump (RCP) operation.

First, there are small break LOCAs for which delayed RCP trip can result in higher fuel cladding temperatures and a greater extent of zircalloy-water reaction.

Using the conservative evaluation model, analyses for these LOCAs result in a violation of the Emergency Core Cooling System (ECCS) Acceptance Criteria as stated in 10CFR50.46.

The currently approved Westinghouse Evaluation Model for small break LOCAs was used to perform these analyses and found acceptable for use by the NRC in letters 83-10 c and d.

Therefore, to be consistent with the conservative analyses performed, the RCPs should be tripped if indications of a small break LOCA exist.

Secondly, Westinghouse and the Westinghouse Owners'roup have always felt that the RCPs should remain operational for non-LOCA transients and accidents where their operation is beneficial to accident mitigation and recovery.

This position was taken even though a design basis for the plant is a loss of off-site power.

Plant safety is demonstrated in the Final Safety Analysis Reports for all plants for all transients and accidents using the most conservative assumption for reactor coolant pump operation.

In keeping with these two positions, a low RCS pressure (symptom based)

RCP trip criterion was developed that provided an indication to the operator to trip the RCPs for small break LOCA but would not indicate a need to trip the RCP for the more likely non-LOCA transients and accidents where continued RCP operation is desirable.

The basis for this criterion is included in the generic Emergency

Response

Guideline (ERG) Background Document (EW Basic Revision, Appendix A).

Relevant information regarding the expected results of using this RCP trip criterion can be derived from the transients which resulted from the stuck open steam dump valve at North Anna in 1979, the steam generator tube rupture at Prairie Island in 1980 and the steam generator tube rupture at Ginna in 1982.

The RCPs were tripped in all three cases.

However, a study of the North Anna and Prairie Island transients indicated

that RCP trip would not have been needed based on the application of the ERG trip criterion.

The Ginna event, however, indicated a need to review the basis for the RCP trip criterion to allow continued RCP operation for a steam generator tube rupture for low head SI plants.

Thirdly, it has always been the position of Westinghouse and the Westinghouse Owners'roup that if there is a doubt as to what type of transient or accident is in progress, the RCPs should be tripped.

Again, the plants are designed to migate the effects of all transients and accidents even without RCP operation while maintaining a large margin of safety to the public.

The existing emergency operating procedures reflect this design approach.

Lastly, it remains the position of Westinghouse and the Westinghouse Owners'roup that RCP trip can be achieved safely and reliably by the operator when required.

An adequate amount of time exists for operator action for the small break LOCAs of interest.

The operators have been trained on the need for RCP trip and the emergency operating procedures give clear instructions on this matter.

In fact, one of the initial operator activities is to check if indications exist that warrant RCP trip.

Westinghouse and the Westinghouse Owners'roup will undertake a

two-part program to address the requirements of NRC letters 83-10 c and d

based on the aforementioned positions for the purpose of providing more uniform RCP trip criteria and methods of determining those criteria.

In the first part of the program, revised RCP trip criteria will be developed which provides an indication to the operator to trip the RCPs for small break LOCAs requiring such action but will allow continued RCP operation for steam generator tube ruptures, less than or equal to a double-ended tube rupture.

The revised RCP trip criteria will also be evaluated against other non-LOCA transients and accidents where continued RCP operation is desirable in order to demonstrate that a need to trip the RCPs will not be indicated to the operator for the more likely cases.

Since this study is to be utilized for emergency response guideline development, better estimate assumptions will be applied in the consideration of the more likely scenarios.

The first part of the program will be completed and incorporated into Revision 1 of the Emergency

Response

Guidelines developed by Westinghouse for the Westinghouse Owners'roup.

The scheduled date for completion of Revision 1 is July 31, 1983.

The second part of the program is intended to provide the required justification for manual RCP trip.

This part of the program must necessarily be done after the completion of the first part of,the program.

The schedule for completion of the second part of the program is the end of 1983.

The preferred and safest method of pump operation following a small break LOCA is to manually trip the RCPs before significant system voiding occurs.

No attempt will be made in this program to demonstrate the acceptability of continued RCP operation during a small break LOCA.

Further, no request for an exemption to 10CFR50.46 will be made to allow continued RCP operation during a small break LOCA.

II h

e

~

• AI

DETAILED RESPONSE TO NRC LETTER 83-10 C

Each of the requirements stated in the attachment of NRC letter 83-10 c will now be discussed indicating clearly how it will be addressed.

The organization of this section of the report parallels the attachment to NRC letter 83-10 c.

I. Pump Operation Criteria Which Can Result in RCP Trip During Transients and Accidents.

1.

Set pints for RCP Trip The Westinghouse Owners'roup (WOG) response to this section of requirements will be contained in Revision 1 to the Emergency

Response

Guidelines (ERG) scheduled for July 31, 1983.

Carolina Power

& Light Company's plans for incorporation of the WOG ERGs into plant-specific procedures were detailed in CP6L's response to Generic Letter 82-33 forwarded on April 15, 1983.

a)

As stated

above, Westinghouse and the Westinghouse Owners'roup are developing revised RCP trip criteria which will assure that the need to trip the RCPs will be indicated to the operator for LOCAs where RCP trip is considered necessary.

The criteria will also ensure continued forced RCS flow for:

1) steam generator tube rupture (up to the design basis, double-ended tube rupture) 2)

the other more likely non-LOCA transients where forced circulation is desirable (e.g.,

steam line breaks equal to or smaller than one stuck open PORV)

NOTE:

Event diagnosis will not be used.

The criteria developed will be symptom based.

The criteria being considered for RCP trip are:

1)

RCS wide range pressure

< constant 2)

RCS subcooling

< constant 3)

Wide range RCS pressure

< function of secondary pressure Instrument uncertainties will be accounted for.

Environmental uncertainty will be included if appropriate.

No partial or staggered RCP trip schemes will be considered.

Such schemes are unnecessary and increase the requirements for training procedures and decision making by the operator during transients and accidents.

b)

The RCP trip criteria selected will be such that the operator will be instructed to trip the RCPs before voiding occurs at the RCPT 3

H

~

c)

The criteria developed in Xtem la 'above is not expected to lead to RCP trip for the more likely non-LOCA and SGTR transients.

However, since continued RCP operation cannot be guaranteed, the emergency response guidelines provide guidance for the use of alternate methods for depressurization.

d)

The Emergency

Response

Guidelines contain specific guidance for detecting, managing and removing coolant voids that result from flashing.

The symptoms of such a situation are described in these guidelines and in detail in the background document for the guidelines.

Additionally, explicit guidance for operating the plant with a vaporous void in the reactor vessel head is provided in certain cases where such operation is needed.

Carolina Power

& Light Company will include this guidance in the plant-specific procedures for SHNPP and provide operator training on this subject.

e)

The plant-specific procedures being prepared for SHNPP from the WOG ERGs will address the status of Reactor Coolant Pump auxiliaries, provide provisions for the restoration of those auxiliaries, and require tripping of the Reactor Coolant Pumps (RCPs) if those auxiliaries are not available.

Operator training will emphasize the need to trip RCPs if those auxiliaries are not available.

Discussed in la and 1c.

2.

Guidance for Justification of Manual RCP Tri The Westinghouse Owners'roup (WOG) response to this section of requirements will be reported separately at the end of 1983.

Upon receipt of the WOG report, Carolina Power

& Light Company will review the analyses and conclusions in the report and confirm the applicability of the results to the Shearon Harris Nuclear Power Plant.

A schedule for review of the report will be forwarded upon receipt of the report.

a)

A significant number of analyses have been performed by Westinghouse for the Westinghouse Owners'roup using the currently approved Westinghouse Appendix K Evaluation Model for small break LOCA.

This Evaluation Model uses the WFLASH Code.

These analyses demonstrate for small break LOCAs oY concern, if the RCPs are tripped two minutes following the onset of reactor conditions corresponding to the RCP trip setpoint, the predicted transient is nearly identical to those presented in the Safety Analysis Reports for all Westinghouse plants.

Thus, the Safety Analysis Reports for all plants demonstrate compliance with requirement 2a.

The analyses performed for the Westinghouse Owners'roup will be used to demonstrate the validity of this approach.

b)

Better estimate analyses will be performed for a limiting Westinghouse designed plant using the WFLASH computer code with

J N

II E

better estimate assumptions.

These analyses will be used to determine the minimum time available for operator action for a range of break sizes such that the ECCS acceptance criteria of 10CFR50.46 are not exceeded.

It is expected that the minimum time available for manual RCP trip will exceed the guidance contained in N660.

This will justify manual RCP trip for all plants.

3.

Other Considerations a)

A schedule for addressing the qualification of the instrumentation used to sense parameters applicable to the RCP trip setpoint will be provided upon receipt of the WOG report at the end of 1983.

b)

The Emergency

Response

Guidelines contain guidance for the timely restart of the reactor coolant pumps when conditions which will support safe pump start-up and operation are established.

This guidance will be incorporated into the plant-specific procedures and operator training.

c)

The SHNPP operator training program will instruct operators with respect to their responsibility in tripping RCPs.

That training,.

will include prioritization of actions following engineered safety features actuation.

II. Pump Operation Criteria Which Will Not Result in RCP Trip During Transient and Accidents.

The preferred and safest method of operation following a small break LOCA is to manually trip the RCPs.

Therefore, there is no need to address the criteria contained in this section.

(6684JJSmf)

0 J

P