ML20199D458

Forwards Preliminary Review Results & Comments Re Responses to Generic Ltr 85-12 Concerning Implementation of TMI Action Item II.K.3.5, Automatic Trip Reactor Coolant Pumps. Alternatives for Completing Review Listed

ML20199D458
Person / Time
Site:	Mcguire, Catawba, McGuire, 05000000
Issue date:	06/10/1986
From:	Youngblood B Office of Nuclear Reactor Regulation
To:	Tucker H DUKE POWER CO.
References
TASK-2.K.3.05, TASK-TM GL-85-12, TAC-49653, TAC-53290, NUDOCS 8606200329
Download: ML20199D458 (17)

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c  %, UNITE 3 8TATES g a NUCLEAR REGULATORY COMMISSION s wAsHmGTON, D. C. 20666 10 JUN 1986 .

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Docket Nos.: 50-369, 50-370,

. 50-413, 50-414 l

Mr. H. B. Tucker, Vice President Nuclear Production Department

, Duke Power Company

, 422 South Church Street Charlotte, North Carolina 28242

Dear Mr. . Tucker:

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Subject:

TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps, Catawba and McGuire Nuclear Stations, Units 1 and 2 The NRC staff is reviewing your responses to Generic Letter 85-12 for Catawba 7 and McGuire Nuclear Stations regarding implementation of TMI Action Item II.K.3.5,

" Automatic Trip of Reactor Coolant Pumps." The enclosure describes our pre-liminary review results and contains several coments reflecting a need for clarifications and additional infomation. Although the enclosure is based upon McGuire, it also substantially applies to Catawba.

We recomend the following course of action with respect to completion of this review effort:

(1) That the enclosure serve as the basis for a telephone conference call prior to initiation of preparation of a response.

(2) That the NRC staff document results of the telephone conference, including our understanding of your responses, and forward to you.

(3) That you confirm the accuracy of our description of the responses, identify any appropriate changes, and provide any supplemental information needed to complete the review.

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Upon satisfactory completion of the above, the staff would complete its SERs.

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e60620o329 860610 9 PDR ADOCK OLOO P

N Mr. H. B. Tucker i}

An alternate approach for Item (1) above would be a meeting. An alternate to Items 1 through 3 would be your treatment of the enclosure as a fomal request j for additional information. In any event, because of the substantial similarity F.

between the McGuire and Catawba approaches for RCP trip, we recomend you con-sider a combined comunication covering both facilities.

1 Contact your Project Manager, Darl Hood at (301) 492-8060 or Kahtan Jabbour at j (301)492-7367 to arrange a conference call or meeting, to schedule your reply dates, or if you have questions regarding the enclosure. l

} Sincerely, I h M

! $ B. J. Youngblood, Director PWR Project Directorate #4 Division of PWR Licensing-A, NRR

Enclosure:

As stated 7

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Mr. H. B.. Tucker I O JUN 1386 s

T An alternate approach for Item (1) above would be a meeting. An alternate to Items- 1 through 3 would be your treatment of the enclosure as a formal request for additional:infomation. In any event, because of the substantial similarity

. between the McGuire and Catawba approaches for RCP trip, we recomend you con-sider a combined comunication covering both facilities, e Contact you'r Project Manager, Darl Hood at (301) 492-8060 or Kahtan Jabbour at (301)492-7367 to arrange a conference call or meeting, to schedule your reply dates, or if you have questions regarding the enclosure.

Sincerely,

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B. J. Youngblood, Director PWR Project Directorate #4 Division of PWR Licensing-A, NRR

Enclosure:

As stated

. cc: See next page DISTRIBUTION:

b A L PDR NSIC PRC System PWRf4 R/F MDuncan DHood KJabbour BJYoungblood R/F EJordan BGrimes JPartlow

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OELD WLyon CBerlinger bW /* W bL PWRf4:DPWR-A PWRf4:DPWR-A PWRg4hPWR-A c PWRf4:D )WR-A DHood:kab- KJabbour MDdhcan # BJYoungblood 04/ 3t/86 3

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, Mr. H. B. Tucker Duke Power Company McGuire Nuclear Station CC:

Mr. A. Carr Dr. John M. Barry Duke Power Company Department of Environmental Health

,- P. O. Box 33189 Mecklenburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Carolina 28242 Charlotte, North Carolina 28203 1

,i Mr. F. J.-Twogood County Manager of Mecklenburg County l Power Systems Division 720 East Fourth Street

! Westinghouse Electric Corp. Charlotte, North Carolina 28202 j P. O. Box 355 Pittsburgh, Pennsylvania 15230 Chainnan, North Carolina Utilities Commission 5 Mr. Robert Gill Dobbs Building Duke Power Company 430 North Salisbury Street l Nuclear Production Department Raleigh, North Carolina 27602 P. O. Box 33189 Charlotte, North Carolina 28242 Mr. Dayne H. Brown, Chief

. Radiation Protection Branch J. Michael McGarry, III, Esq. Division of Facility Services I Bishop Libennan, Cook, Purcell Department of Human Resources

)

and Reynolds P.O. Box 12200 -

) 1200 Seventeenth Street, N.W. Raleigh, North Carolina 27605 Washington, D. C. 20036 Senior Resident Inspector c/o U.S. Nuclear Regulatory Connission

Route 4, Box 529 Hunterville, North Carolina 28078 .

i Regional Administrator, Region II l U.S. Nuclear Regulatory Commission, 3

101 Marietta Street, N.W., Suite 2900

. Atlanta, Georgia 30323 3- .

L. L. Williams Operating Plants Projects Regional Manager l Westin house Electric Corporation - R&D 701 P. O. ox 2728 Pittsburgh, Pennsylvania 15230

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- Mr. H. B. Tucker t

. Duke Power Company Catawba Nuclear Station cc:

William L. Porter, Esq. North Carolina Electric Membership Duke Power Company ' Corp.

P.O. Box 33189 3333 North Boulevard Charlotte, North Carnlina 28242 P.O. Box 27306

Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.

Bishop, Liberman, Cook, Purcell Saluda River Electric Cooperative, and Reynolds Inc.

1200 Seventeenth Street, N.W. P.O. Box 929 Washi'ngton, D. C. 20036 Laurens, South Carolina 29360 North Carolina MPA-1 Senior Resident Inspector

. Suite 600 Route 2, Box 179N 3100 Smoketree Ct. York, South Carolina 29745 P.O. Box 29513

[ Raleigh, North Carolina 27626-0513 Regional Administrator, Region II U.S. Nuclear Regulatory Comission, Mr. C. D. Markham 101 Marietta Street, NW, Suite 2900 Power Systems Division Atlanta, Georgia 30323 t

f Westinghouse Electric Corp.

P.O. Box 355

! Pittsburgh, Pennsylvania 15230 Robert Guild,'Esq.

2759 Rosewood Drive i NUS Corporation Columbia, South Carolina 29205 2536 Countryside Boulevard i Clearwater, Florida 33515 Palmetto Alliance

! 2759 Rosewood Drive l Mr. Jesse L. Riley, President Columbia, South Carolina 29205 Carolina Environmental Study Group .

I 854 Henley Place Karen E. Long Charlotte, North Carolina 28208 Assistant Attorney General N.C. Department of Justice Richard P. Wilson, Esq. P.O. Box 629 4 Assistant Attorney General Raleigh, North Carolina 27602 S.C. Attorney General's Office

.J P.O. Box 11549 Spence Perry, Esquire

Columbia, South Carolina 29211 Associate General Counsel Federal Emergency Management Agency Piedmont Municipal Power Agency Room 840 100 Memorial Drive 500 C Street

[ Greer, South Carolina 29651 Washington, D. C. 20472 Mark S. Calvert, Esq. Mr. Michael Hirsch Bishop, Liberman, Cook, Federal Emergency Management Agency Purcell & Reynolds Office of the General Counsel

. 1200 17th Street, N.W. Room 840 Washington, D. C. 20036 500 C Street S.W. .

Washington, D. C. 20472 Brian P. Cassidy, Regional Counsel Federal Emergency Management Agency, Region !

J, W. McCormach POCH Boston, Massachusetts 02109 ,

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Catawba (amendments) cc:

Mr. Heyward G. Shealy, Chief Bureau of Radiological Health South Carolina. Department of Health and Environmental Control 2600.Bu11 Street-Columbia, South Carolina 29201 County Manager of York County York County Courthouse York. South Carolina 29745

. Attorney General P.O. Box 11549 Columbia, South Carolina 29211 T

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ENCLOSURE

. PRELIMINARY REVI.EW OF.

l REACTOR COOLANT PUMP (RCP) TRIP CRITERIA MCGUIRE UNITS 1 AND 2 RCP TRIP

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INTRODUCTION Listed.below ars the evaluation criteria the staff is applying to RCP trip review and our~ preliminary appraisal of whether sufficient information was

! provided for.the staff to complete its review. The organization of thi's g Enclosure is a staff provided guidance or evaluation criterion statement based-E upon Gener'ic Letter 85-12 (Ref. 1), generally'followed by a staff critique of

the initial _ licensee submittal (Ref. 2).

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The staff plan to complete the review is to use this Enclosure as the basis for a telephone conference call prior to initiation of a significant response

-effort on the part of the licensee. The staff then plans to document the

< results of- the preliminary review and the telephone conference call, and to provide that documentation to the licensee. Then the licensee can provide a 1 written response regarding the accuracy of the staff perception of the telephone conference call, together with any desired corrections and i supplemental information. This should allow the staff to complete the review f and prepare a Safety Evaluation Repo'rt (SER). An alternate, if the licensee desires (and which is r4 presently' contemplated), is for the staff to prepare l"

w i formal-questions 4 r ' ice see consideration.

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N The submittal for McGuire is similar to that for Catawba (Ref. 3). Most of the staff response to Reference 3 is identical to that.of Reference 2. Since i many of the same points apply, the staff buggests the reviews be coordinated i to minimize time requirements, imposed upon both Duke and the staff.

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OVERALL GUIDANCE PERTINENT TO RCP TRIP During a-small break accident in certain' break size ranges, there exists a window in time during which tripping RCPs will make the accident worse.

Therefore, in a small break situation, one must trip RCPs prior to entering the window. If one wishes to depend upon manual trip,'two criteria are applicable:

1. -One must show that at least 2 minutes exist within which to trip RCPs-following " receipt of a trip signal" using licensing calculations as a I basis.

2. One must show that at least 10 minutes exist within which to trip RCPs following " receipt of a trip signal" using best estimate calculations as a basis.

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If, for some reason, the RCPs have.not been tripped within 10 minutes of the time at which plant conditions indicate trip should be performed, they are to be left running until after the window is closed. Closure can be indicated by parameters such as regaining both adequate subcooling margin and pressurizer

level after they have been lost.

Analyses are required to establish timing relative to items 1 and 2, as well as to establish the' dimensions of the window.

-It-is desirable to leave pumps running for control purposes during other transients and accidents, including steam generator tube rupture accidents of l sizes up to one tube broken. Therefore, insofar as is practical, procedures

[ and criteria should be developed to attain this goal. Note that leaving pumps

.I running during "non-break" transients and accidents is not a 100% requirement, O as cuntrasted to the small break, where trip must be accomplished to remain in I compliance with the regulations. (Failure to trip as required could lead to i exceeding Appendix K specified temperatures.) For "non-break" transients and accidents, RCPs may be tripped when desirable. If in doubt, the small break i ~

criteria are to be applied.

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New plants coming on line should have dealt with RCP trip prior to power

,. operation.

Note much of the work pertinent to the above criteria has been done on a generic basis, and is applicable to individual plants., Where this is the' case, it is sufficient to establish applicability, and the generic work need not be repeated on a plant specific basis.

SPECIFIC EVALUATION CRITERIA AND COMENTS The evaluation criteria are generally those provided in Reference 1, including the Safety Evaluation and its appendices, which were an enclosure to Reference 1.

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A. Determination of RCP Trip Criteria i

Demonstrate and justify that proposed RCP-trip setpoints are adequate for small-break LOCAs but will not cause RCP trip for other non-LOCA transients and accidents such as SGTRs. This is to include performance

?- of safety analyses to prove the adequacy of the setpoints.

Consider using partial or staggered RCP-trip schemes.

Staff Evaluation. Duke has selected Reactor Coolant System (RCS) subcooling as the criterion for tripping the Reactor Coolant Pumps (RCP).

The subcooling option was previously identified by the staff as the j second choice of the three options described by the Westinghouse Owners f Group (WOG) in their evaluation of RCP trip. It is also a choice that i few other plants are making. This is an acceptable selection if sufficiently justified. Therefore, the staff requests information pertinent to this selection.

The quantitative value associated with subcooling that is to be used as the trip criterion is not identified. Are the RCPs to be tripped on loss L

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of subcooling or of subcooling margin and, if the latter, what is the margin?

.A1. Identify the instrumentation to be used to determine the RCP trip set point, including the degree of redundance of each parameter signal needed for the criterion chosen. Establish the quality level for the instrumentation, identify the basis for the sensing-instruments' design features, and identify the basis for the degree of redundance.

Staff Evaluation. Duke has elected to use subcooling as determined by

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the plant computer as the primary indicator of the need to trip the RCPs.

If the plant computer is not available, they will use instrumentation readings available to the operator on control room indicators, and will apply graphs containing pressure and temperature information to determine

! the subcooling.

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p Input to the plant computer consists of wide and low range RCS pressure,

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j , a hot leg temperature from each of the hot legs, and core exit thermocouples. Low range pressure is used, if valid, since this is stated to have a smaller uncertainty than wide range. If unavailable, then wide range pressure is used. (The wording is such that the staff is 1

not clear on whether wide range pressure is the same as pressurizer l pressure.) If this is unavailable, then pressurizer pressure is used.

l The highest five valid thermocouple readings are used to determine an average. This average is compared with the four loop wide range temperatures, and the highest valid indication from this comparison is

used as the temperature input.

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I The quality level and bases for these pressures and temperatures are not identified, nor are the specific instruments. In the case of the core exit thermocouples, the identity is obvious, and no further information t

t is necessary. Are the hot leg temperatures RTD readings from manifolds which are connected to the hot legs? Where are the pressure i determinations made (i.e. , are they pressurizer connections or

[ connections to some other part of the RCS pressure boundary)? -

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Yhat constitutes a " valid" reading and how is validity determined? How is the influence of adverse conditions factored into the decisions?

A2. Identify the instrumentation uncertainties for both normal and adverse containment conditions. Describe the basis for the selection of the adverse containment parameters. Address, as appropriate, local conditions, such as fluid jets or pipe whip, which might influence instrumentation reliability.

Staff Evaluation. Instrumentation uncertainty for normal and adverse environmental conditions is listed as follows:

Instrument Normal Adverse Conditions Conditions p RCS Wide Range Pressure 63 213

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RCS Low Range Pressure 29 125 Loop Wide Range Hot Leg Temperature, F 8.1 8.5 Core Exit Thermocouple Temperature, F 7.2 8.5 where the core exit temperature error increases with temperature, the error shown is for 530 F, and this error is assumed applicable below 530 F. ,

A value of 3 psig .is used as the basis for selection of adverse 4 containment parameters. This approximately corresponds to the high-high containment pressure setpoint.

Local conditions which might influence instrument reliability are stated as addressed in the McGuire response to NUREG-0588. It would be helpful if this were summarized with respect to RCP trip, including consideration of conditions outside of containment that may influence un. certainty. For example, has Duke surveyed the wiring and connections between the pressure transmitters and the control room to assure that a steam line 5

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• Q break inside or outside of containment will not introduce problems with the pressure readings in the control room?

Are any of the pressure transmitters located outside containment, and, if so, what is the instrument response time?

! A3. In addressing criterion selection, provide consideration of uncertainties associated with the WOG supplied analyses values. These uncertainties are to include uncertainties in computer program results and uncertainties resulting from plant specific features not representative of the generic data group.

If a licensee (or applicant) determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip, it is recommended that a f .

more discriminating plant-specific procedure be developed. Licensees (or l f' applicants) should take credit for all equipment (instrumentation) f~

available to the operators for which the licensee (or applicant) has sufficient confidence that it will be operable during the expected

conditions.

Staff Evaluation. Calculations of instrument uncertainties are

. summarized, and comparisons are discussed between plant data and calculations in the Westingliouse Owners Group (WOG) information. Duke states that the calculated overall uncertainty for McGuire is from -10 F I to +10 F for the subcooling trip point.

The licensed Westinghouse LOFTRAN computer code is referenced for l

1 performance of the non-LOCA analyses. The computer program result uncertainties evaluation is based on the assumption of no changes-in initial plant conditions (such as full power, pressurizer level, all Safety Injection (SI) pumps running,-and all Auxiliary Feed Water (AFW) pumps running). The major contributors to uncertainty are stated to be

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. break flow rate, SI flow rate, decay heat generation rate, and AFW flow h rate. Parametric studies are summarized in which the major uncertainties

[ are stated to be due to the break flow model and SI flow inputs.

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l~' Duke has not directly addressed such topics as the accuracy of the i numerical solution scheme or of nodalization. Further, there is no determination of the influence of equipment or operational failures.

Information pertinent to the former result from comparisons of the LOFTRAN code to operational and experimental data, and as a result will

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! have been included in the uncertainty number. Determination of equipment

.or operational failures is not a necessity as long as the expected configuration of the plant is addressed since the objective of RCP trip is to provide reasonable assurance of not tripping for transients for which a trip is undesirable. It is not necessary to establish that one l will never trip since the plant is capable of being safely controlled if an unnecessary trip does occur. Thus, no additional information is needed for the staff to complete review of this item.

. B. Potential Reactor Coolant Pump Problems T

Bl. Assure that containment isolation, including inadvertent isolation, will

. not cause problems if it occurs for non-LOCA transients and accidents.

Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-LOCA situation is confirmed to prevent seal damage or failure. Confirm that' containment isolation with continued pump operation will not lead to seal or pump damage.or failure.

Staff Evaluation. The Duke response is essentially that either seal injection or thermal barrier cooling will continue regardless of the status of containment isolation, and therefore, there is not problem with h seal damage. Component Cooling Water (CCW) and service water are

,i . isolated on Phase B isolation (high-high containment pressure). RCPs are tripped promptly under these conditions.

The staff requests additional information pertinent to these points. For j example, do the conditions which lead to containment isolation result in termination of any portion of the Chemical and Volume Control System 1 (CVCS) and is there an indirect effect upon RCP seal injection? If CCW I

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pump operation is terminated, how long does it take for the pump to be restarted and flow restored to the thermal barrier heat exchangers and other RCP associated components? What are the implications? Information should be provided pertinent to restart of RCPs following restoration of services leading to a trip. Items such as trip parameters, operator response and timing of operations should be identified. The staff does not need a large volume of material on these topics, but does need a brief mention with perhaps reference to procedures in a list provided with the submittal.

82. Identify the components required to trip the RCPs, including relays, power supplies and breakers. Assure that RCP trip, when necessary, will occur. Exclude extended RCP_ operation in a voided system where pump head is more .than 10% degraded unless analyses or tests can justify pump and

. pump-seal integrity when operating in voided systems. If necessary, as a f result of the location of any critical component, include the effects of

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H adverse containment conditions on RCP trip reliability. Describe the basis for the adverse containment parameters selected.

Staff Evaluation. The major ~ components associated with RCP trip are identified, as is their location.

. A brief consideration should be given to the potential for adverse conditions outside containment and the implications, if any. For example, can a steam line break outside of containment introduce difficulties with respect to the equipment of interest here?

The timing of operations associated with alternate operator actions required to trip the RCPs should be mentioned. For example, if the operator attempts a trip from the control room and fails, how long will it take to trip from an alternate location, including travel time? (See Item C1, below.)

RCP operation in a voided system is not mentioned. -

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l C. Operator Training and Procedures (RCP Trip)

C1. Describe the operator training program for RCP trip. Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running. Also cover priorities for actions after engineered safety features actuation. -

Assure that training and procedures provide direction for use of individual steam generators with and without operating RCPs.

Assume manual RCP trip does not occur earlier than two minutes after the RCP-trip set point is reached.

i Determine the time available to the operator to trip the RCPs for the limiting cases if manual RCP trip is proposed. Best Estimate p . calculational procedures should be used. Most probable plant conditions

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should be identified and justified by the licensee, although NRC will accept conservative estimates in the absence of justifiable most probable conditions.

Justify that the time available to trip the RCPs is acceptable if it is less than the Draft ANSI Standard N660. If this is the case, then address the consequences if RCP trip is delayed. Also develop contingency procedures and make them available for the operator to uss in l case the RCPs are not tripped in the preferred time frame.

Staff Evaluation. A discussion of training is presented, but the background philosophy is not clearly described, and some of the above points are not addressed. For example, what is the general need to trip RCPs as contrasted to keeping them running? What is the RCP restart philosophy? If one has tripped RCPs, and desires to restart them for control purposes and to aid in plant cooldown, what requirements must be met?

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. C2. Identify those procedures which include RCP trip related operation:

(a) RCP trip using WOG alternate criteria (b) RCP restart (c) Decay heat removal-by natural circulation (d) Primary system void removal (e) Use of steam generators with and without RCPs operating (f) RCP trip for other reasons Ensure that emergency operating procedures exist for the timely restart of the RCPs when conditions warrant.

f Staff Evaluation. Duke has presented a listing of selected procedures and provides discussions which address these topics.

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l REFERENCES -

1. Thompson, Hugh L. Jr. , " Implementation of TMI Action Item II.K.3.5,

' Automatic Trip of Reactor Coolant Pumps' (Generic Letter No. 85-12)",

NRC Letter Addressed to All Applicants and Licensees with Westinghouse g ) Designed Nuclear Steam Supply Systems (NSSSs), Jun. 28, 1985.

2. Tucker, Hal B., "McGuire Nuclear Station, Docket Nos. 50-377 and 50-370, Generic Letter 85-12," Letter to Hugh L. Thompson, Jr. , NRC, from Duke Power Company, August 22, 1985.

3. Tucker, Hal B., " Catawba Nuclear Station, Docket Nos. 50-413 and 50-414, Generic Letter 85-12," Letter to Hugh L. Thompson, Jr. , NRC, from Duke Power Company, August 22, 1985.

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