ML18019A803

Forwards Addl Info to Util 850903 Response to TMI Item II.K.3.5 Concerning Reactor Coolant Pump Trip Criteria,Per 860305 Request.Info Should Close Out SER Confirmatory Item 33(b)

ML18019A803
Person / Time
Site:	Harris
Issue date:	05/01/1986
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
TASK-2.K.3.05, TASK-TM NLS-86-106, NUDOCS 8605060225
Download: ML18019A803 (16)

Text

REQULA Y INFORMATION DISTRIBUTIO YBTEM (RXDS)

ACCESSION NBR: 86050b0225 DOC. DATE: 86/05/01 NOTARIZED:

NO DOCKET N FACIL:50-400 Shearon Harris Nuclear Power Planti Unit ii Car olina 05000400 AUTH. NAME AUTHOR AFFILIATION ZIMMERMANiS. R.

Carelina Power 5 Light Co.

REC IP. NAME REC IP XENT AFFILIATION DENTONp H, R Office of Nuclear Reactor Regulationi Direc tor (p ost 851125

SUBJECT:

Forwards addi info to util 850903 response TMI Action Item II. K. 3. 5 concerning reactor coolant pump trip criteria'er 860305 request. Info should closeout BER Confirmatory Item (b).

DISTRIBUTION CODE:

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TITLE: Licensing Submittal:

PSAR/FSAR Amdts 8c Related Correspondence NOTES: *pp 1 ication for permi t renewal fi led.

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Carolina Power 8 Light Company MAY SERIAL: NLS-86-106 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO.

1 - DOCKET NO.50-000 REACTOR COOLANT PUMP TRIP CRITERIA

Dear Mr. Denton:

Carolina Power k Light Company (CPRL) hereby submits additional information in response to TMI Action Item II.K.3.5 concerning reactor coolant pump trip criteria.

The attached information is submitted in response to informal questions from the NRC reviewer',based on CPRL's previous letter of September 3, 1985.

Based on a March 5,

- 1986 telephone conversation with your staff, the attached information should be sufficient to close out the Shearon Harris Nuclear Power Plant Safety Evaluation Report Confirmatory Item 33(b).

If you have any additional questions on this subject, please call.

Yours very truly, 3HE/ccc (3622NLU)

Attachment cc:

Mr. B. C. Buckley (NRC)

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

Dr. 3. Nelson Grace (NRC-RII)

Mr. Travis Payne (KUDZU)

Mr. Daniel F. Read (CHANGE/ELP)

Wake County Public Library S.

Zim er man ager Nuclear Licensing Section Mr. Wells Eddleman Mr. 3ohn D. Runkle Dr. Richard D. Wilson Mr. G; O. Bright (ASLB)

Dr-. 3. H. Carpenter (ASLB)

Mr. 3. L. Kelley (ASLB)

Bb050b0225 Sb0501 PDR ADOClt, 0500OO00 E

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

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Al.

For instrumentation identified in your response, provide a discussion of quality level, environmental qualification, and redundancy.

RESPONSE

The instrumentation used to determine the RCP trip setpoint are RCS Wide Range Pressure Transmitters PT-000 and PT-001.

These instruments are redundant.

These instruments are not Class lE, but are powered from an Uninterruptible Power Source.

Two additional backup indications of RCS pressure are provided by transmitters Nos. PT-002 and PT-003.

These transmitters are environmentally and seismically qualified. They are redundant and Class lE. Instruments PT-002 and PT-003 are used as backup indications due to their slow response times caused by their location outside containment.

(3622NLU/+<)

A2.

Address the local conditions which might permanently damage instrumentation, II such as fluid jets and pipe whip. Also address the operator response to instruments under abnormal conditions when one instrument may be inoperative and the other providing information with a large uncertainty.

RESPONSE

As discussed above, the SHNPP RCP trip parameter is RCS wide range pressure, instruments PT-000 and PT-OOI, which provide redundant indication at the main control board.

There are also two transmitters, PT-002 and PT-003, located outside containment and, therefore, not subject to adverse containment conditions and associate instrument uncertainties; these also provide redundant indication at the main control board.

Local conditions such as fluid jets and pipe whip can permanently damage either transmitter PT-000 or FT-001.

Due to pipe routing and instrument location, there are no fluid jets or pipe whip which could per manently damage both transmitters.

Transmitter PT-000 can be permanently damaged by a hot leg break (Loop 3) or a break in its own impulse line. Transmitter PT-001'can be permanently damaged by a CVCS or hot leg break (Loop 1) or a break in its own impulse line.

Per SHNPP Procedure OMM-OOI, "Operation-Conduct of Operations," operators are required to believe instrumentation unless and until the readings are proven invalid. There are three cases to consider under conditions of RCS depressurization with abnormal containment conditions:

1)

V/ith one of the two instruments fully inoperable and reading off-scale,'the operator would verify actual RCS pressure using the redundant pressure instrument and would take appropriate actions for RCP trip based on this redundant instrument.

2)

V/ith the affected instrument providing inaccurate but conservative pressure readings (i.e., lower than actual pressure but not off-scale), the operator under guidance of OMM-001 would be required to believe this indication and would trip the RCPs.

(3622NLU/m < )

3)

With the affected instrument providing inaccurate but non-conservative pressur e readings (i.e., higher than actual pressure but not off-scale), the operator under guidance of OMM-001 would be required to believe this indication and would use the redundant instrument to identify when RCP tr ip criteria are met, since the faulted instrument would indicate a pressure higher than the RCP trip criteria.

In all three cases, with one of the two redundant instruments damaged, timely tripping of the RCPs is accomplished.

(3622NLU/mt )

II

A3.

Discuss the computer analyses and the ability of the selected parameter (RCS wide range pressure) to differentiate between accidents.

Discuss uncertainties associated with computer program results.

Provide a quantitative basis for the selection of a specific parameter.

Provide the methodology for differentiating between normal operating conditions and abnormal conditions in selection of RCP trip value.

RESPONSE

The analyses performed by the Westinghouse Owners'roup (WOG) to calculate limits for RCP trip criteria based on RCS dynamic response under non-LOCA events is documented in the Emergency Response Guidelines (ERG) Background Document, Revision I, "RCP Trip/Restart," dated September I, 1983.

This document (part of the ERG Revision 1 package which has been submitted to the NRC) is the basis for all SHNFP Emergency Operating Procedures and is in use throughout the industry as the basis document for emergency procedures development for Westinghouse PV/Rs.

The technical information in Revision I is identical to the original revision which was accepted by the NRC, "Supplemental SER on WOG ERG."

The analyses performed by SHNPP to calculate plant-specific RCP trip parameters, as well as the documentation of uncertainties associated with these calculations, was submitted in response to Item A3 in CPRL's previous response dated September 3, 1985. Allplant-specific calculations were manually performed; there is no computer analyses nor any associated computer uncertainties.

Since all three calculated parameters are less than the limiting values determined by the V/OG analysis, all three are acceptable to differentiate between a'OCA situation and a non-LOCA (especially SGTR) situation.

There is, therefore, no quantitative basis for selection of RCS wide range pressure as the RCP trip parameter since all three parameters could be so justified. The basis for our selection, which was an II engineering and human factors judgment, was documented in CPRL's previous response.

Operator recognition of adverse containment conditions is a covered item in RO and SRO license training. The SHNPP criteria for adverse containment conditions are consistent with or conservative to the WOG ERGs Revision 1 (see Executive

(>622NLU/mf)

Volume, Generic Instrumentation section).

Any of the following conditions are considered adverse for the purpose of selecting the appropriate value of the RCP trip parameter (or any other parameter in the EOPs which has both a normal and adverse condition value):

1)

Containment Pressure above High-1 setpoint of 3 PSIG 2)

Containment Radiation Level above 10 R/HR 3.

Containment Radiation Integrated Exposure above 10 RADS (determined by engineering analyses)

(3622NLU/+< >

Bl.

Discuss RCP restart following other difficulties which result in RCP trip, such as loss of CCW or RCP parameters exceeding trip limits; identify these limits.

RESPONSE

When restart of RCPs is addressed in EOPs and the RCPs have experienced other problems unrelated to the accident which resulted in RCP trip (i.e., not based on LOCA RCP trip parameter), operation of RCPs is in accordance with procedures AOP-018 "Abnormal RCP Operation" and OP-100 "Reactor Coolant System."

An exception to this is procedure EOP-FRP-Cl "Response to Inadequate Core Cooling," in which RCP restart is required to establish core cooling when conditions have degraded to a point beyond plant design basis.

In such a case, the procedure does not require normal prerequisites to RCP start.

This deviation is in accordance with WOG ERGs Revision 1.

RCP trip criteria under non-emergency conditions are addressed in AOP-018 "Abnormal RCP Operation" and are in accordance with Westinghouse guidance provided in the RCP technical manual.

The RCP is required to be tripped under any of the following conditions:

1)

Any bearing temperature exceeds 190 F 2)

Seal inlet/pump bearing water temperature exceeds 230'F 3)

Motor winding temperatures exceed 300'F 0)

Seal injection flow is lost and either of the following occurs:

a)

CCW flow to thermal barrier heat exchanger below 00 GPM b)

CCW Hx outlet temperature exceeds 105'F (or 130'F with RCS temperature below 000'F) 5)

Loss of CCW flow to RCP motor oil coolers for more than 10 minutes.

(3622NLU/el )

6)

Vibration levels exceed any of the following:

a) b)

c) d)

20 MILS shaft 5 MILS frame 15 MILS shaft and increasing at 1.0 MIL/HRor more 3 MILS frame and increasing at 0.2 MIL/HRor more (3622NLU/mf )

B2.

The applicant should determine whether components required for RCP trip are located where they willbe affected by high energy line breaks, or other environments which may result in abnormal conditions in the the Reactor Auxiliary Building.

RESPONSE

As discussed in CPRL's previous response dated September 3, 1986, the following components are required for RCP trip:

MCB Control Switch 6.9kV Auxiliary Bus Breaker RCP Trip Coil No.

1 125Y DC Control Power 102SA Bus A, Cub 5 TC1-102 DP-IA-SA 100SA Bus B, Cub 9 TC1-100 DP-1A-SA 106SA Bus C, Cub 2 TC1-106 DP-1A-SA These components are located in the Reactor Auxiliary Building, Elevation 305 ft and 286 ft. There are no high energy lines or sources of adverse environments located near this equipment which could affect the operation of this equipment.

(3622NLU/mf)

Cl.

Provide detail of operator training and the understanding of the need to trip RCPs as contrasted to keeping them running.

RESPONSE

RO and SRO license training is in accordance with the IVOG ERGs Revision 1 and WOG Loss of Reactor or Secondary Coolant Training Program, 3une 1985, RCP Trip Criteria Lesson Plan.

Plant-specific lesson plans were developed using these documents as references.

These lesson plans provide instruction on the four subject areas documented and listed in our previous submittal; license candidates also receive training concerning the desirability to run RCPs during,non-LOCA events versus the requirement for tripping RCPs under small-break LOCA conditions.

For nearly all analyzed accident scenarios, operation of RCPs is either beneficial or neutral to mitigating the consequences of the accident.

Under conditions of a small break in the RCS cold leg (SBLOCA), however, operation of the RCPs beyond a specific time period and subsequent shutdown occurring in a specific time interval can result in more severe core uncovery and damage than results from

.analyzed FSAR accidents.

Operation of RCPs under SBLOCA conditions causes the liquid inventory of the RCS to be forced out the break at a higher rate than would occur if the RCPs were not operating.

This is due to forced flow through the RCS maintaining liquid in the loop piping; this increases the time during which liquid is escaping the break rather than steam. If the RCPs continued to operate throughout the event, core cooling is ensured despite excessive RCS inventory depletion due to forced steam flow through the core and RCS depressurization continues until the safety injection flow becomes sufficient to restore RCS inventory. However, if the RCPs trip during a specific time interval which varies with break size and decay heat generation rate, the excessive loss of liquid caused by RCP operation results in core uncovery to a depth and for a time sufficient to cause peak cladding temperatures greater than those which result from other analyzed accidents.

(Core uncovery occurs due to phase separation upon loss of forced RCS flow.) The beginning of this interval corresponds to the point at which the break would uncover and steam rather than liquid would escape the RCS if the RCPs were not operating.

Steam release is desirable because the higher specific energy of steam compared to liquid water (3622NLU/m f )

results in greater heat removal from the RCS, allowing depressurization of the RCS and increased injection flows to recover core liquid level. This situation can only occur when the liquid inventory of the RCS is such that the loops become voided; the steam generator V-tubes must become steam-filled, and this occurs when RCS pressure at the top of the V-tubes decreases to saturation conditions. If RCP trip is delayed, more liquid leaves the RCS and additional steam forms.

When RCP trip occurs, and the steam-liquid separation occurs, the core uncovers to a greater depth.

Further, since there is more steam in the RCS, a longer period of time is required for it to escape the break and depressurize the RCS.

During this time, the core is uncovered until the RCS depressurization results in sufficiently increased injection flow to recover core level.

Therefore, although RCPs should be operated for non-LOCA events to provide normal pressurizer spray and forced RCS flow (especially for SGTR events),

tripping of the RCPs must be accomplished under SBLOCA conditions prior to the point when continued operation of RCPs cause excessive depletion of RCS inventory such that the break would be uncovered if the RCPs were not operating, allowing steam to escape through the break.

Tripping of the RCPs is accomplished at the m'ain control board using the control switch for each pump.

In the event of malfunction, the affected pump(s) would be locally tripped at the breaker; these are located in the switchgear rooms in the Reactor Auxiliary Building (elevation 286 ft.) one level below the control room and accessible via the stairwell located in the control room area.

No special keys or tools are required to access and locally trip the breakers.

(3622NLU/mf )

C2.

Discuss RCP restart under emergency operating procedures.

RESPONSE

RCP restart under emergency conditions is in accordance with WOG ERGs Revision 1. The following SHNPP EOPs address RCP restart:

SHNPP ERGs TITLE EPP-000 ES-0.1 EPP-005 ES-0.2 EPP-006 ES-0.3 EPP-007 ES-0.0 EPP-008 EPP-009 EPP-015 ES-1.1 ES-I.2 ECA-2.1 EPP-020, ECA-3.1, Reactor Trip Response Natural Circulation Cooldown Natural Circulation Cooldown with Steam Void (With RVLIS)

Natural Circulation Cooldown With Steam Void (Without RVLIS)

SI Termination Post-LOCA Cooldown and Depressurization Uncontrolled Depressurization of all Steam Generators SGTR With Loss of Reactor Coolant:

EPP-021 EPP-022 FRP-Cl FRP-13 ECA-3.2 ECA-3.3 FR-C.1 FR-I.3 Subcooled Recovery SGTR With Loss of Reactor Coolant:

Saturated Recovery SGTR Without Pressurizer Pressure Control Response to Inadequate Core Cooling

Response

to Voids in Reactor Vessel (3622NLU/III( )

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