NRC Generic Letter 1985-12

NRC Generic Letter 1985-012: Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps

ML031150698
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill, Crane
Issue date:	06/28/1985
From:	Thompson H Office of Nuclear Reactor Regulation
To:
References
GL-85-012, NUDOCS 8507010252
Download: ML031150698 (22)

IIIW~&1985 TO ALL APPLICANTS AND LICENSEES WITH WESTINGHOUSE (W) DESIGNED NUCLEAR STEAM

SUPPLY SYSTEMS (NSSSs)

SUBJECT: IMPLEMENTATION OF TMI ACTION ITEM II.K.3.5, "AUTOMATIC TRIP OF

REACTOR COOLANT PUMPS" (GENERIC LETTER NO. 85-12 )

Gentlemen:

The purpose of this letter is to inform you of (1) the staff's conclusions regarding the Westinghouse Owners Group (WOG) submittals on reactor coolant pump trip in response to Generic Letters 83-lOc and d, and (2) provide guidance concerning implementation of the reactor coolant pump trip criteria.

Our Safety Evaluation (SE) on this subject is enclosed for your use.

With regard to the WOG submittals referenced in Section V of the enclosed SE,

we conclude that the methods employed by the WOG to Justify manual reactor coolant pump (RCP) trip are consistent with the guidelines and criteria provided in Generic Letters 83-lOc and d. The approved Westinghouse Small Break LOCA Evaluation Model was used to demonstrate compliance with

10 CFR 50.46 and Appendix K to 10 CFR Part 50.

We have determined that the information provided by the WOG in support of the alternative RCP trip criteria is acceptable on a generic basis. A suitable reactor coolant pump trip criterion can be selected by each licensee to minimize reactor coolant pump trip during steam generator tube ruptures and non-LOCA events, while still providing for RCP trip for small break LOCAs.

With regard to implementation, we note that the WOG RCP trip methodology allows applicants/licensees to select among three alternate RCP trip criteria.

The selection is based upon obtaining maximum discrimination between a small break LOCA (which requires RCP trip) and a steam generator tube rupture (which does not require RCP trip). In reviewing the WOG RCP trip criteria, we note that the process of criterion selection involves a number of considerations which were assigned plant-specific status by the WOG during the process of the trip criteria review.

Accordingly, we request that operating reactor licensees select and implement an appropriate RCP trip criterion based upon the WOG methodology. Schedules for submittal of information requested in Section IV of the SE (refer to Appendix A for considerations associated with Generic Letters83-10c and d)

should be developed with your individual project managers within 45 days from a receipt of this letter. The requested information does not constitute a n

CONTACT

D. Jaffe X28140 1-.

VII

I-

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requirement but only identifies information specified in Generic Letters83-10c and d which has not been provided under the WOG generic program. In the event that licensees decide not to trip the RCP (an option provided for in Generic Letters83-10c and d), they should respond to the questions in Section IV of the SE and refer to Appendix B of the SE. Applicants should provide the appropriate response to the extent that this information is known at this time.

Those applicants and licensees who choose not to endorse the WOG methodology should submit a schedule for submittal of plant specific RCP trip criteria or Justification for non-trip of RCPs within 45 days of receipt of this letter.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0065 which expires September 30, 1985.

Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.

If you believe further clarification regarding this issue is necessary or desirable, please contact Mr. D. Jaffe (301 492-8140).

Sincerely, Original Signed by Hugh L.Thompson, Jr.

Hugh L. Thompson, Jr., Director Division of Licensing Enclosure:

SafetX Evaluation cc w/enclosure:

Service Lists Distribution:

DJaffe Memo FIle PKreutzer GLainas ORPMs BSheron Retyped by CSchum 06/27 DL: AD/SA*

DCrutchfield

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06/26/85 06/26/85 06/26/85 06/26/85 06/26/85 l/4/185

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information specified in Generic Letters 83-lOc and d which has not been provided under the WOG generic program. In the event that licensees decide not to trip the RCP (an option provided for in Generic Letters83-10c and d), they should respond to the questions in Section IV of the SE and refer to Appendix B of the SE. Applicants should provide the appropriate response to the extent that this information is known at this time.

Those applicants and licensees who choose not to endorse the WOG methodology should submit a schedule for submittal of plant specific RCP trip criteria or justification for non-trip of RCPs within 30 days.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0065 which expires September 30, 1985.

Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.

If you believe further clarification regarding this issue is necessary or desirable, please contact Mr. D. Jaffe (301 492-8140).

Sincerely, Hugh L. Thompson, Jr., Director Division of Licensing Enclosure:

Safety Evaluation cc w/enclosure:

Service Lists Distribution:

DJaffe Memo FIle PKreutzer GLainas ORPMs BSheron Retyped by CSchum 06/27 DL i (p DC field

• SEE PREVIOUS PAGE FOR CONCURRENCE & /11/85 ORB#3:DL* ORB#3:DL* C-RSB* ORB#3:DL* AD:OR:DL* D: DL

PKreutzer DJaffe BSheron EButcher GCLainas HThompson

06/26/85 06/26/85 06/26/85 06/26/85 06/26/85 / /85

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Generic Letters83-10c and decide not to trip the RCP (an option provided for in of the SE and refer to d), they should respond to the questions in Section IV

should provide the appropriate response to Appendix B of the SE. Applicants the extent that this information is known at this time.

the WOG methodology Those applicants and licensees who choose not to endorse RCP trip criteria or should submit a schedule for submittal of plant specific justification for non-trip of RCPs within 30 days.

of Management and This request for information was approved by the OfficeSeptember 30, 1985.

Budget under clearance number 3150-0065 which expiresto the Office of Comments on burden and duplication may be directed New Executive Office Management and Budget, Reports Management, Room 3208, Building, Washington, D.C. 20503.

issue is necessary or If you believe further clarification regarding this desirable, please contact Dr. B. Sheron (301 492-7460).

Sincerely, Hugh L. Thompson, Jr., Director Division of Licensing Enclosure:

Safety Evaluation cc w/enclosure:

Service Lists Distribution:

DJaffe Memo FIle PKreutzer GLainas ORPMs BSheron C-RSB A DL D:DL

ORB#3:DL E t9her nas HThompson PK~rezer eJaffe BSheron

54 //t 85 6 /z1/85 I2X/85 / /85

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

/

REGULATION

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR

FOR THE

WESTINGHOUSE WNERS GROUP

REACTOR COOLANT PUMP TRIP

I. INTRODUCTION

all licensees to TMI Action Plan Item II.K.3.5 of NUREG-0737 required consider other solutions to the small-break loss-of-coolant-accident pumps (RCPs) was not (LOCA) problems because tripping the reactor coolant of the RCPs in the case of considered the ideal solution. Automatic trip solution was found. A

a small-break LOCA was recommended until a better NRC programs concerning RCP

summary of both the industry programs and theb, c, d, and e, which are trip is provided in Generic Letters 83-lOa, W. J. Dircks to the NRC

included in the NRC report, SECY-82-475, from Coolant Trip Issue"

Commissioners, "Staff Resolution of the Reactor the NRC guidelines and (November 30, 1982). SECY-82-475 also provided II.K.3.5, "Automatic criteria for the resolution of TMI Action Item Trip of Reactor Coolant Pumps."

pump trip In SECY-82-475 the NRC concluded: "...that appropriate would not require RCP

setpoints can be developed by the industry thatforced convection circulation trip for those transients and accidents where aid to the operators, yet and pressurizer pressure control is a major for those small LOCAs where would alert the operators to trip the RCPs in core damage."

continued operation or delayed trip might result in the enclosures SECY-82-475 also stated: "The resolution provided that for whatever mode of (Generic Letters 83-10 is intended to ensure technical basis for that pump operation a licensee elects, a) a soundmeet the Commission's rules and decision exists, b) the plant continues to remain running for those regulations, and c) as a minimum, the pumps willconvection cooling and non-LOCA transients and accidents where forced control. This would pressurizer pressure control would enhance plant to approximately the (SGTR) up include steam generator tube ruptures design basis event (one tube)."

reports to the NRC in The Westinghouse Owners Group (WOG) submitted two 83-lOc and d. The response to the Westinghouse specific Generic Letters, RCP Trip Criteria"

first report provided an "Evaluation of Alternate of Manual (Reference 1). The second report provided the "Justification

2). The WOG also pro- RCP Trip for Small Break LOCA Events" (Reference to our request for vided additional information (Reference 3) in response WOG submittals. We have also this information, based on the review of the our review (Reference 4).

performed analyses of selected events to support

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Appendix A to this report summarizesSection I of the enclosure to Generic Letter 83-10 for "Pump-Operation Criteria that Can Result in RCP Trip During Transients and Accident," and Appendix B summarizesSection II,

"Pump-Operation Criteria That Will Not Result in RCP Trip During Transients and Accident."

II. SUMMARY

The WOG has developed a set of three alternative reactor coolant pump (RCP) trip criteria - each one being reported to be equally suitable in meeting the intent of Generic Letter 83-10. The revised criteria replace the current RCP trip criterion of low reactor coolant system (RCS)

pressure, which could result in RCP trip for SGTR and non-LOCA events.

The objective of the WOG study was to evaluate alternative RCP trip criteria to determine if a criterion could be established to reduce the probability of RCP trip for SGTRs and non-LOCA events, while still providing for RCP trip for small break LOCAs.

The parameters which were considered for the evaluation of alternative RCP

trip criteria included RCS pressure, RCS subcooling, and secondary pressure dependent RCS pressure (RCS/secondary pressure differential).

Because SGTRs and most non-LOCA events will not result in adverse containment conditions, the normal instrument uncertainties associated with the measurement of these parameters can be utilized in evaluating the effectiveness of these alternative criteria in preventing pump trip for most SGTRs and non-LOCAs.

The alternative RCP trip criteria which were evaluated are:

1. RCS pressure with normal instrument uncertainties.

This criterion would be established in the same manner as the current RCS pressure criterion, with the exception that the normal instrument uncertainties would be utilized in determining the RCP trip setpoint for normal containment conditions. The instrument uncertainties associated with post-accident containment conditions would continue to be used to determine the RCP trip setpoint for adverse containment conditions.

2. Reactor coolant subcooling.

This method would provide a direct indication of the need for pump trip, since pump trip is not required as long as the reactor coolant remains subcooled. The RCP trip setpoint would be established as zero degrees subcooling in the RCS hot legs, plus the uncertainty in the subcooling monitor to assure that the pumps are tripped before subcooling is actually lost. The normal instrument uncertainties

-would be used for normal containment conditions, whereas the instru- ment uncertainties associated with post-accident containment conditions would be used for adverse containment conditions.

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3. Secondary pressure dependent RCS pressure.

With the current method of using RCS pressure, the trip criteria is conservatively derived assuming that the secondary pressure is at the lowest secondary safety valve set pressure. However, the secondary pressure may actually be significantly less than this value, particu- larly if the condenser steam dump system is in operation. With this method, the RCS pressure setpoint for pump trip would be continuously evaluated based on the actual secondary pressure. The alternate RCP

trip criterion can also be expressed as the RCS/secondary pressure differential. The combined instrument uncertainties for the RCS and secondary pressure measurements would be included in determining the RCP trip setpoint. The normal instrument uncertainties would be used for normal containment conditions, whereas the instrument uncertainties associated with post-accident containment conditions would be used for adverse containment conditions.

The results of the small break LOCA analysis demonstrate that each of the alternative criteria is adequate in providing an indication for the operator to trip the RCPs for small break LOCAs. Thus, the selection of the RCP trip criterion can be based on the capability to preclude a pump trip for SGTRs and non-LOCAs. The minimum values of each of the parameters used to evaluate the alternative criteria were also determined for SGTRs and non-LOCA events for each category of plants in the study. A methodology was provided to determine the RCP trip setpoints for each of the three alternative criteria. Using the results of the evaluation and the setpoints calculated for each of the alternative criteria, each utility can determine which of the criteria will prevent pump trip for SGTRs and non-LOCA events for their respective plants. iThe criterion that is considered most appropriate in providing pump trip discrimination between LOCAs and SGTR

or non-LOCA events can then be selected by the utility for each plant.

Based on these studies, the WOG concluded that the RCP trip criterion can generally be implemented using existing qualified instrumentation already available in the plants, and additional instrumentation is not required.

The WOG followed the guidelines provided in Generic Letter 83-lOc and d to Justify manual RCP trip for small-break LOCAs. (See Appendix A,

Section D.) The WOG studies concluded that:

1. Every Westinghouse plant's FSAR ECCS analysis demonstrates compliance with 10 CFR 50.46 if operator action to trip the RCPs is taken within two minutes after the RCP trip criterion is reached.

2. Most probable best estimate analyses indicate that in all Westinghouse plants the RCPs may be tripped at any time during a small break LOCA event without reaching clad temperatures of 2200'F.

The highest PCT calculated with most probable best estimate assumptions was 1255 0F.

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The WOG concluded that automatic reactor coolant pump trip is not required since adequate time for manually tripping the RCPs is demonstrated using

10 CFR Part 50, Appendix K assumptions as well as most probable best estimate analysis results. It was also concluded that the most probable best estimate analysis results demonstrate that the RCPs can be tripped at any time during the LOCA (if the operator should fail to trip the pumps when the trip criterion is reached) without incurring unacceptable clad temperature results. Therefore, the WOG concludes that the existing guidelines in Revision 1 of the Emergency Response Guidelines (ERGs) are sufficient and complete with respect to RCP status for all accident situations, and no additional "missed RCP trip setpoint" steps are required.

The methods (References 5 and 6) employed by the WOG to Justify manual RCP

trip are consistent with the guidelines and criteria provided in Generic Letters 83-1Oc and d.

We have reviewed thk assumptions and models employed by the WOG to study steam generator tube rupture (SGTR) and non-LOCA events. The LOFTRAN

computer program (Reference 7) was used to study these events. Best estimate assumptions and models were used. However, the SGTR break flow model incorporated into LOFTRAN does result in higher than expected mass flow rates for a given break size. (This model was approved for SAR SGTR

analyses, where the high flow rates result in a conservative evaluation of offsite dose.) The WOG position with respect to the use of this conservative model is that the analysis results are bounding for the design basis SGTR event of a single tube.

The WOG considered all other FSAR Chapter 15 non-LOCA events for evaluation against the alternate RCP trip criteria. It was concluded that the feedline and steamline breaks needed to be considered because their transient characteristics would be the most limiting with respect to the three criteria. For the steamline break accident, a "credible" (4.5 inch diameter) break size was considered - essentially equivalent in size to a single steam generator PORV failing open. For the feedline break, a full double-ended rupture of the main feedwater line was considered.

We have reviewed the assumptions, models and plant groupings used to perform the SGTR and non-LOCA studies and have determined that the information provided is acceptable. Table 1 provides a summary of the WOG

studies. We believe that the three RCP trip criteria may be marginal for some plants for the SGTR event. We base our conclusion on the following:

1. The SGTR event gives the minimal values for all three alternative RCP

trip criteria for all but a few plants, and

2. The uncertainty analysis of instrument error provided by the WOG for use to evaluate the trip set points for each alternative criteria (for both normal and adverse containment conditions) may not be bounding for all plants.

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In particular, the reactor coolant system pressure set point RCP trip criterion appears to offer the least in reducing unnecessary RCP trip.

This confirms our position as discussed in SECY-82-475 and Generic Letter 83-10.

The WOG objective for the SGTR and non-LOCA analyses was to consider design basis accidents with more realistic assumptions, to enable the development of a RCP trip criterion which would provide reasonable assurance of continued pump operation for these accidents. While it is possible that other accident conditions could result in more limiting parameter values, the design basis accidents which were defined for the analyses combined with the conservatisms which are incorporated in the analytical model provide assurance that the analysis results will be bounding for most SGTR and non-LOCA events. The WOG does not consider it to be practical or necessary to develop a RCP trip criterion which will provide for continued pump operation for all possible SGTR and non-LOCA events. It would not be a safety problem if RCP trip should occur for a SGTR or non-LOCA event, since the plant safety systems are designed to handle those accidents with a loss of offsite power and, therefore, with RCP

trip. The objective was to demonstrate that the RCPs will remain on for most of the expected cases of these accidents, so that the operator can retain normal pressurizer pressure control and will not be required to open the pressurizer PORVs. In addition, maintaining forced reactor coolant system flow will reduce the likelihood of generating voids in the reactor vessel upper head region.

The WOG response to our concern that none of the three alternative criteria would prevent RCP trip for a SGTR or a non-LOCA event (on a plant specific bases) is the recommendation to use the criterion which demonstrates the greatest discrimination capability.

In doing so, the WOG expects that a large range of SGTRs and non-LOCA events still would not require RCP trip. In the event of RCP trip occurring for SGTRs and non-LOCAs, the WOG position is that the Emergency Response Guidelines (ERGs) provide specific contingency actions to recover the plant even though RCP operation is not available. Also, specific RCP

restart steps are built into the ERGs where deemed beneficial although they are not required for safe plant shutdown. The WOG expects, however, that at least one of the alternative criteria will be successful in preventing pump trip for SGTRs and non-LOCA events for each of the plants.

The studies performed by the WOG to determine the transient characteristics for the SGTR and non-LOCA events were based on best estimate input assumptions and models (to the extent practical with the computer programs used). Based on our experiences, with other thermal-hydraulic programs used to perform similar types of analyses, we believe there are uncertainties associated with the numerical results of any calculated system transient. Each licensee must consider these uncertainties when selecting the criterion which demonstrates the greatest discrimination capability, and be prepared to explain how they were considered during future inspections.

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The generic nature of the WOG submittals concerning RCP trip by nature do not include any plant specific information, other than that needed to determine plant groupings for analysis. We have therefore included a section (Implementation) in this report which describes those plant specific items we require each licensee to address when Incorporating the RCP trip criteria into the plant procedures.

III CONCLUSIONS

We have determined that the information provided by the WOG for the Justification of manual reactor coolant pump trip is acceptable. The methods employed by the WOG to justify manual reactor coolant pump trip are consistent with the guidelines and criteria provided in Generic Letters83-10c and d. The approved Westinghouse Small Break LOCA Evaluation Model was used to demonstrate compliance with 10 CFR 50.46 and Appendix K to

10 CFR Part 50.

We have determined that the information provided by the WOG in support of the alternative reactor coolant pump trip criteria is acceptable. A

suitable reactor coolant pump trip criterion can be selected by each licensee to minimize reactor coolant pump trip during steam generator tube ruptures and non-LOCA events, while still providing for RCP trip for small break LOCAs.

The results presented by the WOG, for the plant groups studied, imply that one of the alternative RCP criteria would prevent RCP trip for the design basis SGTR and for design basis non-LOCA events. This would be a true statement if the numerical results from the calculation performed were error free and if each plant responds exactly as the simulation model predicts. Also, the uncertainty analysis for instrument error would have to be bounding for each plant, with normal containment conditions.

Adverse containment conditions are not expected for design basis SGTRs or non-LOCA events.

We believe the analysis tools employed by the WOG are capable of qualitatively providing the appropriate information to evaluate the alternate RCP criteria. It should be obvious however that the quantitative values provided cannot be considered absolute. In our judgement, the alternate RCP trip criteria, as defined, may provide only marginal assurance of preventing RCP trip for the design base SGTR event for some Westinghouse plants.

We have concluded that the WOG has developed acceptable criteria for tripping the reactor coolant pumps during small-break LOCAs and to minimize reactor coolant pump trip for SGTR and non-LOCA events.

IV IMPLEMENTATION

The generic information presented by the WOG does not address plant specific concerns about instrumentation uncertainties, potential reactor coolant pump problems and operator training and procedures as requested in Generic Letter 83-10. Appendix A contains a summary related to these issues and may be used as a guideline to assure that these issues are adequately addressed.

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In order to complete the response to Generic Letters 83-lOc and d, each W licensee is required to submit the following information to the NRC

Tor plant specific reviews:

A. Determination of RCP Trip Criteria

1. Identify the instrumentation to be used to determine the RCP

trip set point, including the degree of redundancy of each parameter signal needed for the criterion chosen.

2. 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 the instrumentation reliability.

3. In addressing the selection of the criterion, consideration to uncertainties associated with the WOG supplied analyses values must be provided. These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant specific features not representative of the generic data group.

If a licensee determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip, it is recommended that a more discriminating plant-specific procedure be developed. For example, use of the NRC-required inadequate-core- cooling instrumentation may be useful to indicate the need for RCP trip. Licensees should take credit for all equipment (instrumentation) available to the operators for which the licensee has sufficient confidence that it will be operable during the expected conditions.

B. Potential Reactor Coolant Pump Problems

1. Assure that containment isolation, including inadvertent isolation, will not cause problems if it occurs for non-LOCA

transients and accidents.

a. 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.

b. Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.

2. Identify the components required to trip the RCPs, including relays, power supplies and breakers. Assure that RCP trip, when determined to be necessary, will occur. If necessary, as a result of the location of any critical component, include the effects of adverse containment conditions on RCP trip reliability.

Describe the basis for the adverse containment parameters selected.

- 8- C. Operator Training and Procedures (RCP Trip)

1. 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.

2. Identify those procedures which include RCP trip related operations:

(a) RCP trip using WOG alternate criteria Mb) 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

V REFERENCES

1. Westinghouse Owners Group, Letter OG-110, "Evaluation of Alternate RCP

Trip Criteria," December 1, 1983.

2. Westinghouse Owners Group, Letter OG-117, "Justification of Manual RCP

Trip for SBLOCA Events," March 9, 1984.

3. Westinghouse Owners Group, Letter OG-137, "Response to NRC Question on RCP

Trip," October 25, 1984.

4. Lime, J. F., "TRAC Analysis of Small-Break and Tube-Rupture Accidents for the Evaluation of Westinghouse Alternate Reactor-Coolant-Pump Trip Criteria," Los Alamos National Laboratory, February 1985 (DRAFT).

5. Esposito, V., et al., "WFLASH - A FORTRAN IV Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP-8200 Rev. 2, June

1984.

6. Bordelon, F. M., et al., "LOCTA - IV Program: Loss-of-Coolant Transient Anaylsis," WCAP-8301, June 1984.

7. Burnett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907-P-A,

April 1984.

Attachments:

Appendix A

Appendix B

TABLE I Sl'flHART TABLE or voc NEACTQR COOLANT PUMP TRIP SM 9IS

51 ACC. UP S.C. S.C. NO MINIMUM ICS MINIMUM RtS MIN RCSISECONDARY

POWER CUT PUMP C1A HEAD VALVE TUBE LOAD PRESSURE SUSCOOLINE Dair. PRESSURE

NO. DOCKET OR TAC LEVEL orr WEIR "I pIES TEOP SET PT I.D. TATE ---- (PSIA) ---- ----- (F) ---- ---- (PSI)-

ELANT LOOr NO. NO. (IWT) HEAD tI/ TiN (PSIUC RIC (PSIA) (IN) (F) SCTR SLI ILU lSUR SLR rLe SCm SLI fLB

VOGTLE I 4 50-424 0 3411 Hi Y N 400 C 12O0 .406 557 1753 1774 1164 56 71 4t 415 811 817 VOGTLE 2 4 50-425 0 3411 HI i N I40 C 120

SEABROOK I 4 50-443 0 3411 Rl i N 405 C 1206 SEABROOK I 4 50-444 0 341t H iT N t00 C 1200

MILLSTONE 3 4 50-423 0 3411 N t00 C 1200

WOLF CREEK 4 50-482 0 3411 N 40l C 1200

CALLAVlA 4 50-483 0 3411 HliT N to0 C 1200 (

BYRON I 4 50-454 0 3411 HI I N 400 C 1200 .444 557 146 1774 1418 57 71 4? 44 Sit 817 BYRON I 4 56-455 0 3411 HI Y N 400 C 1200

BRAIDVOOD I 4 50-456 0 3411 Nl T N 400 C 1205 IRAIDWOOD I 4 56-457 0 3411 R iT N t00 C 1205 Mc GUIRE I 4 50-341 49 is3 3411 H iT I 400 C 10OE

nC tUIRE 2 4 50-370 0 3411 HI T I 401 C 1200

CATAVIA I 4 50-413 0 3411 HI 7 I 400 C 1200

CATAV0A 2 4 55-414 0 3411 HR r I 405 C 1200

MARBLE HILL I 4 50-546 0 3411 H iT N too C 1200

MARBLE RILL 2 4 50-547 6 3411 Hi T N t00 C 1200

WATTS BAR I 4 .50-390 0 3411 Hi 7 I 460 C 1200

WATTS BAR 2 4 51-391 0 3411 HI T I 405 C 1200

COMANCHE PEAK 1 4 50-445 0 3411 Hi T I 400 C 1206 COMANCHE PEAK I 4 50-444 0 3411 NI r N 4o C 1200

TROJAN 4 50-344 47477 3301 Hi t V N 400 H 1100 .775 557 1524 1774 1t4e 52 71 4? 535 111 817 (

ZION I 4 50-213 441113 3250 NI b N N 400 N 1100 .775 547 1407 1774 1141 ll 71 49 404 ill I17 ZION 2 4 50-304 47614 3250 HI N N t00 H 1101 DIABLO CANTON 2 4 50-323 0 3411 NI I N 400 R 1100 .775 547 1473 1774 13IJ 57 71 0 5614 111 117 SALEM I 4 50-272 4I174 3411 HI I N d00 H 1106 SALEM 2 4 50-311 40470 3411 Ni I N 466 R 1100

SEOUOTAH I 4 50-327 47944 3411 HI N T 406 c 119l SEGUOYAN 2 4 50-321 51321 3411 Ill N T 400 C 1100

TABLE I (CON'T) ASUMART TABLE Of VODREACTOR COOLANT PUMP TRIP STUDIES

51 ACC. UP S.C. S.C. NO MINIMHU RCS MINIMNU RCS 111t RCSISECONDARY

POVER CUT FUM? CAS HEAD VALV TUBE LOAD FRESSURE SUICOOLIN; DIET. PRESSURE

NO. DOCKET OR TAC LEVEL OFF VEIR URI PRES TEt? SET PT I.D. TATE ----(PSIA) ---- .. iS ()- ----1P5I- - -

PLANT LOOP No. NO. (IYT) READ tiN TIN (FSI)C HIC (PSIA) (IN) (fl SCUT SLI 1l3 SGTR SLI fIl SGTR SLI FlU

D C. COOK I 4 50-315 49701 3250 Hi N N I00 H 1100 .775 547 1443 1774 1011 55 71 41 542 111 117 B.C. COOK I 4 50-311 49702 3391 Hi N N I05 H 1100

DIABLO CANYON I 4 50-275 41454 3335 Hi Y N 400 R 1100

SOUTH TE2AS I 4 5.-495 0 3500 IN N 400 H 1300 .444 517 1407 1530 1551 40 44 I2 453 414 557 SOUTH TEIAS 2 4 50-499 0 3800 IN N 400 H 1300

INOIAN POINT 2 4 50-247 41472 2758 LO N 45t H 1100 .775 547 1110 1155 1157 31 10 1 21931050 154

(

INDIAN POINT 3 4 50-2ll 41457 3025 LO N t00 H 1100 .775 547 1211 1155 1557 32 90 10 315 1050 154 SUMMER 3 50-375 41115 2775 Hi N 400 C 1200 .444 557 1543 1431 1123 SI 70 71 541 711 522 SHERON HARRIS I 3 50-400 0 2775 Hi N 100 C 1200

SHERON HARRIS 2 3 50-401 0 2775 Hi N 605 C 1206 FARLEY I 3 50-348 41145 2152 NI N 501 R 1100 .775 547 1234 1436 1723 37 70 71 350 710 122 FARLEY 2 3 50-314 4141V 2452 HI N 400 H 1100

NORTH ANNA I 3 50-331 41445 2775 Hi N 400 R 1100

NORTH ANNA 2 3 50-337 40144 2775 Hi N 4l6 H 1100

SURR7 I 3 50-250 41481 2441 HI N N 400 R 1100

SURRY 2 3 50.281 41412 2441 Hi N N t0 N 1100

DEAVER VALLEY I 3 50-334 49415 2452 HI N 400 U floe BEAVER VALLEY 2 3 50-412 0 2l5t Hi N 101 R 1100 .775 547 1147 1434 1123 31 70 71 271 710 522 (

H.R. ROIINSON 3 50-241 404l0 2321 LO ' T N !1P a 1100 .775 547 1147 1741 I'IM 31 52 10 301 1150 102 TURKEY POINT 3 3 30-250 49479 200 L 10 N 40a R 1101 TURKEY POINT 4 3 50-251 41460 2200 LO T N 410 N 1100

PRAIRIE ISLAND I 2 50-i22 41443 1450 IN N N 700 a 1100 .775 547 1341 1117 1751 31 85 It 351 1100 1047 PRAIRIE ISLAND I 2 50-304 41444 1450 IN N N 711 R 1100

I 50-30S 40411 14S0 IN N N 700 R 1100 .775 547 1253 1777 1751 31 15 14 341 110t 1147 R.E.CINMA 2 50-144 4101 1310 LO T N 700 R 1100 .775 II? 11t7 1111 I123 21 74 1 1305 440 t00

POINT BEACH I 2 S0-244 40454 1520 O I N 100 H 1100

POINT BEACH 2 I 50-301 49457 1520 O I N 700 R 1100

APPENDIX A

PUMP-OPERATION CRITERIA THAT CAN RESULT IN RCP TRIP

DURING TRANSIENTS AND ACCIDENTS

A. The NRC staff has concluded that if sufficient time exists, manual action is acceptable for tripping the RCPs following a LOCA provided certain conditions are satisfied.

B. Potential problem areas should be considered in developing RCP-trip setpoints and methods.

1. Tripping RCPs causes loss of pressurizer sprays.

a. This produces a need to use PORVs in some plants to control primary pressure.

b. PORVs have frequently failed to close.

c. Despite testing, PORV operational reliability has not improved significantly.

2. Tripping RCPs tends to produce a stagnant region of hot coolant in the reactor-vessel upper elevations.

a. Hot stagnant coolant has flashed and partially voided the upper vessel region during depressurization or cooldown operational events.

b. Operators are not completely familiar with the significance of an upper-head steam bubble.

c. Operators have difficulty controlling coolant conditions to avoid or control flashing.

d. Operators may take precipitous actions when a steam bubble exists.

3. After tripping the RCPs, decay-heat removal by natural circulation is required. This procedure is used less frequently than controlling with the RCPs and it places more demand on the operators to control the primary-system conditions.

C. Consider the following guidelines in developing RCP-trip setpoints.

1. 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.

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I

a. Assure that RCP trip will occur for all primary-coolant losses in which RCP trip is considered necessary.

b. Assure that RCP trip will not occur for SGTRs up to and includ- ing the design-basis SGTR.

c. Assure that RCP trip will not occur for other non-LOCA tran- sients where it is not considered necessary.

d. Perform safety analyses to prove that a, b, and c above are achieved.

e. Consider using partial or staggered RCP-trip schemes.

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

g. Assure that symptoms and signals differentiate between LOCAs and other transients.

h. (Westinghouse plant specific) RCP trip is expected to occur for the design-basis SGTR for some Westinghouse plants that have SI

pumps with lower shutoff heads. The exact rupture size above which RCP trip would be required has not been determined.

(1) NRC informed Westinghouse that RCPs should not be tripped for SGTR events such as that which occurred at Ginna, which was essentially equivalent to a design-basis SGTR.

(2) NRC informed Westinghouse that methods should be examined for either improving the RCP-trip setpoints or modifying the plants so that RCPs need not be tripped for design-basis SGTRs.

(3) Restart permission was granted for Ginna with the requirement that supplementary guidelines be developed for RCP trip to assure RCPs would not be tripped for the design-basis SGTR.

(4) NRC agrees with Westinghouse that in the long term using the reactor-vessel-liquid-inventory system to help determine when to do a RCP trip will increase the probability of maintaining RCP operation during non-LOCAs.

i. (Westinghouse plant specific) NRC has concluded that recent information by Westinghouse about wide-range pressure uncertainty indicates that analyses confirming Westinghouse's conclusions about RCP trip setpoints for high-head-SI plants are probably necessary.

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2. 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.

3. Avoid challenges to the PORVs where possible.

a. If setpoints lead to RCP trip even though it is neither required nor desirable for transients or accidents with offsite power available, assure that challenges to the PORVs are avoided that would normally be handled by using pressurizer sprays.

b. Challenges to PORVs could be eliminated by using heated auxiliary pressurizer sprays from a source other than the RCP

discharge.

c. If submittal recommends use of PORVs to depressurize, then licensees need to develop a program for upgrading the PORVs'

operational reliability.

4. Establish guidelines and procedures for cases where RCP trip can lead to hot, stagnant fluid regions at primary-system high points.

a. Describe symptoms of primary-system voiding caused by flashing of hot, stagnant fluid regions including effects on the pressurizer.

b. Specify guidance for detecting, managing and removing the voids.

c. Train operators concerning the significance of primary-system voids for both non-LOCA and LOCA conditions.

5. Assure that containment isolation will not cause problems if it occurs for non-LOCA transients and accidents.

a. Demonstrate that, if water services needed for RCP operation are terminated, they can be restored fast enough once a non-LOCA

situation is confirmed to prevent seal damage or failure.

b. Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.

6. RCP-trip decision parameters should provide unambiguous indicators that a LOCA has occurred and the NRC-required inadequate-core-cooling instrumentation should be used where useful in indicating the need for a RCP trip.

7. NRC recommends that the licensee use event trees to systematically evaluate their set points to minimize the potential for undesirable consequences because of a misdiagnosed event.

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a. Evaluate set points for events with RCP trip when it is preferable the RCPs remain operational.

b. Evaluate set points for events where early RCP trip does not occur and a delayed trip may lead to undesirable consequences.

D. NRC's guidance for justification of manual RCP trip in the-licensee submittals is summarized in this section. This guidance had two purposes.

It was intended to assist plants that can and should rely on manual trip to justify it, and it was also intended to help identify those few plants that may not be able to rely on manual trip.

1. Analyses should demonstrate that the limits set forth in 10 CFR 50.46 are not exceeded for the limiting small-break size and location using the RCP-trip set points developed with the guidance of part C above.

a. Assume manual RCP trip does not occur earlier than 2 minutes after the RCP-trip set point is reached.

b. Include allowances for instrument error.

c. Generic analyses are acceptable if they are shown to bound the plant-specific evaluations.

2. Determine the time available to the operator to trip the RCPs for the limiting cases if manual RCP trip is proposed.

a. Perform the analysis for the limiting small-break size and location identified in D.1 above.

b. Use the most probable best-estimate analysis to determine the time available to trip the RCPs following the time when the RCP-trip signal occurs.

c. Most probable plant conditions should be identified and justi- fied by each licensee.

d. NRC will accept conservative estimates in the absence of justi- fiable most probable plant conditions.

e. Justify that the time available to trip the RCPs is acceptable if it is less than the Draft ANSI Standard N660.

(1) Include an evaluation of operating experience data.

(2) Address the consequences if RCP trip is delayed beyond this time.

(3) Develop contingency procedures and make them available for the operator to use in case the RCPs are not tripped in the preferred time frame.

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(4) No justification is required if the time available to trip the RCPs exceeds the Draft ANSI Standard N660.

E. Assure that good engineering practice has been used for the following areas.

1. Establish the quality level for the instrumentation that will signal the need for RCP trip.

a. Identify the basis for the sensing-instruments' design features chosen.

b. Identify the basis for the sensing-instruments' degree of redundancy.

c. Licensees can take credit for all equipment available to the operators that they have sufficient confidence in its operability during the expected conditions.

2. Ensure that emergency operating procedures exist for the timely restart of the RCPs when conditions warrant.

3. Instruct operators in their responsibility for tripping RCPs for small-break LOCAs including priorities for actions after the engineered safety features actuation occurs.

APPENDIX B

PUMP-OPERATION CRITERIA THAT WILL NOT RESULT IN RCP TRIP

DURING TRANSIENTS AND ACCIDENTS

Consider the following guidelines if the submittal concludes that keeping the RCPs running is both the preferred and safest method of pump operation for small-break LOCAs and other transients and accidents.

A. Evaluate inventory loss.

1. Complete evaluation of LOFT Test L3-6 through the ECCS recovery phase.

2. Evaluate all modeling differences expected between LOFT and a PWR analysis.

B. Evaluate pump integrity.

1. Justify how the pump-seal and pump structural integrity will be assured during extended two-phase flow performance.

2. Include the consequences of pump and/or pump-seal failure in the analyses if their integrity cannot be assured.

3. Include one of the following if continuous RCP operation is expected even with a containment isolation signal.

a. Evaluate the capability to continue RCP operation without essential water services.

b. Evaluate the capability to rapidly restore essential water services.

4. Evaluate the RCP's capability to operate in the accident environment.

5. Evaluate the consequences of RCP failure at any time during the accident if continuous operation in the accident environment cannot be assured.

C. Ensure acceptability of results.

1. Analyses should demonstrate that the 10 CFR 50.46 ECCS acceptance criteria are met with a model in compliance with Appendix K to

10 CFR Part 50.

2. Assume continuous pump operation and also RCP trip at various times if continuous pump operation cannot be assured.

3. NRC will consider a request for an exemption to 10 CFR 50.46 requirements if analyses indicate compliance cannot be achieved.

a. Submittal concludes that compliance with 10 CFR 50.46 would require operating the plant in a less safe condition. This

-2- needs to be supported with a risk/benefit analysis that can take credit for all equipment expected to remain operational during the accident.

b. Submittal concludes that design modifications would not be cost-effective to implement from a safety standpoint.