ML17207A720

Requests Tech Specs for Auxiliary Feedwater Sys Mods & Revision of Normal & Emergency Operating Procedures.Sample Tech Spec Pages Encl

ML17207A720
Person / Time
Site:	Saint Lucie
Issue date:	12/21/1979
From:	Reid R Office of Nuclear Reactor Regulation
To:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
References
NUDOCS 8001080127
Download: ML17207A720 (18)

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Robert E. Uhrfg Vice President Florida Power 8 Light Company Advanced Systems 8 Technol ogy P. 0. Box 529100 Miami, F'lorfda 33152

Dear Mr. Jones:

SUBJEC:

AUTOt@TIC INITIATION OF AUXILIARYFEEDlQTER SYSTEMS AT ST. LUCIE, UNIT NO.

1 In recent communications, your staff has indicated that a proposed design, using control grade components, whfch would automatically initiate the auxf1 fary feedwater systems at your facility upon the loss of main feedwater flow will be submitted in the near future.

This submittal was in response to Short-Term Recormendatfon 2.1.7.a, "Auto Initiation of the Auxiliary Feedwater System",

as clarified in our letter October 30, 1979 which was addressed to all operating nuclear popover plants.

'l<e will review your proposed design against each of the seven positions stipulated fn Short-Term Recoomendatfon 2.1.7.a.

In response to this recormendatfon, some licensees have raised the issue of the applicability of current analysis of a main steam line break or main feedwater line break assuming early initiation of auxiliary, feedwater flow with a failure to limit flev to the affected steam generator.

In question fs whether the change in assumptions would increase the calculated containment pressure or the likelihood of returrr to power.

These questions are believed applicable for either manual or automatic, initiation of the auxiliary feedwater system.

You are requested to resolve this concern by submitting an analysrrs within twenty (20) days after receipt of this letter (telecopfed on date signed).

The enclosure to. this letter provides a list of questions and information you should address as appropriate-As a result of this concern and pursuant to our letter of October 30, 1979, you should not implement automatically initiated AFMS flar until we.have completed our review and issued an approval.

However, to resolve this matter as expeditiously as possible, you should continue with the procurement of equipment and proceed with the installation to the extent possible without activating the automatic-start system or adversely affecting the manual-start AF>lS.

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Docket No. 50-335 Nr. Robert E. Uhrig Vice President Florida Power 6 Light Company Advanced Systems 8 Technology

.P. 0. Box 529100 Niami, Florida 33152

Dear Nr. Uhrig:

SUMECT:

AUTOtNTIC INITIATION OF AUXILIARY FEEDLlATER SYSTEHS AT ST. LUCIE, UNIT NO. I In recent communications, your staff has indicated that a proposed design, using control grade components, which would automatically initiate the auxiliary feedwater systems at your facility upon the loss of main feedwater flow will be submitted in the near future.

This submittal was in response to Short-Term Recommendation 2.1.?.a, "Auto Initiation of the Auxiliary Feedwater System",

as clarified in our letter October 30, 1979 which was addressed to all operating nuclear power plants.

Me will review your proposed design against each of the seven positions stipulated in Short-Term Recommendation 2.1.?.a.

In response to this recomnendation, some licensees have raised the issue of the applicability of current analysis of a main steam line break or main feedwater line break assuming early initiation of auxiliary feedwater flow with a failure to limit f'lax to the affected steam generator.

In question is whether the change in assumptions would increase the calculated contajnment pressure or the likelihood of return to power.

These questions are believed applicable for either manual or automatic initiation of the auxiliary feedwater system.

You are requested to resolve this concern by submitting an analyses or evaluations within twenty (20) days after receipt of this letter (telecopied on date signed).

The enclosure to this letter provided a list of questions and information you should address as appropri ate.

You are requested to propose Technical Specifications for the AFHS modifications.

Sample Technical Specifications are enclosed for your consideration.

In addition, you will need to revise-normal and emergency operating procedures as required by this modification and train the plant operations people as required by these procedures.

Particular attention to the means of controlling the bypass capability of the automatic AFHS turbine start signal is recommended.

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~0 You are requested to propose Technical'pecifications for the AFWS modifications.

Sample Technical Specifications are enclosed for your consideration.

In addition,-

you ivill need to revise normal and emergency operating procedures as required by this modification and train the plant operations people as required by these procedures.

Particular attention to the means of controlling the bypass capability of the automatic AFWS turbine start signal is recormended.

Sincerely,

'a'ZIIines Segno'g I

,Encl osure:

Sample TS Pages cc:

ic/enclosure See next page Robert W. Reid, Chief Operating Reactors Branch 84 Division of Operating Reactors Distribution:

Docket files NRC PDR L PDR TERA JRBuchanan, NSIC NRR Rdg.

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BGrimes" LSaho WGammill RVollmer

-DÃ1ver Merger WRussell RReid HConner RIng ram

Attorney, OELD I8IE (3)

.DEisenhut PErickson TJCarter NFa& ti1>

ACRS (16)

Gray file JRNiller

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NRC FORM 318 {9.76) NRCM 0240 AU,S. GOVERNMENT PRINTING OFFICE: 1979.289 389

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IP fy*y4 Docket No. 50-335 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 OECEMBER e78 Mr. Robert E. Uhrig Vice P resi dent Florida Power 8 Light Company Advanced Systems 8 Technology P. 0.

Box 529100 Miami, Florida 33152

SUBJECT:

AUTOMATIC, INI".IATION OF AUXILIARY FEEDWATER SYSTEMS AT ST. LUCIE, UN.T NO. I In recent comunications, your staff has indicated that a proposed

design, using control grade coriaonents, which would automatically initiate the auxiliary feedwater systems at your facility upon the loss of main feedwater flow will be submitted in the near future.

This submittal was in response to Short-Term RecomIendation 2.1.7.a, "Auto Initiation of the Auxiliary Feedwater Svstem",

as clarified in our letter October 30, 1979 which was addressed to all operating nuclear power plants.

We will review your proposed design against each of the'seven positions stipulated in Short-Tern Recomendation 2.1.7.a.

In response to this recoInIIendation, some licensees have raised the issue of the applicability of current analysis of a main steam line break'r mai n feedwater line break assumi ng early initiation of aux liary feedwater flow with a failure to limit flow to the affected steam 9 nerator.

In question is whether the change in assumptions would increase the calculated containment pressure or the likelihood of return to power.

These questions are believed applicable

'for either manual or automatic initiation of the auxiliary feedwater system.

You are requested to resolve this concern by.submitting an analysis within twenty (20) days after receipt of this letter (telecopied on date signed).

The enclosure to this letter provides a list of questions and information you should address as appropriate.

As a result of this con"em and pursuant to our letter of October 30, 1979, you should not imlement automatically initiated AFWS flow until we have completed our review and issued an approval.

However, to resolve this matter as expeditiously as possible, you should continue with the procurement of equipment and proceed with the installation to the extent possible without activating the automatic-start system or adversely affecting the.-.znual-start'AFWS.

You are requested to propose Technical Specifications for the AFMS modifications.

Sample Technical Specifications are enclosed for your consideration.

In addition, you will need to revise normal and emergency operating procedures as required by this modification and train the plant operations people as required by these procedures.

Particular attention to the means of controlling the bypass capability of the automatic AFWS turbine start signal is recommended.

Sincerely,

Enclosure:

Sample TS Pages cc:

w/enclosure See next page Robert M. Reid, Chief Operating Reactors Branch 44 Division of Operating Reactors

Florida Power 8

ht Company CC:

Robert Lowe ns te in, Esquire Lowenstein,

Newman, Rei s 5 Axel r ad 1025 Connecticut

Avenue, N.W.

Washington, 0.

C, 20036 Norman A. Coll., Esquire McCarthy, Steel, Hector 5 Oavis 14th Floor, First National Bank Building Miami, Florida 33131

~ Mr. Jack Shreve Office of the Public Counsel Room 4, Holland Bldg.

Tallahassee, Florida 32304 Indian River Junior College Library 3209 Virginia Avenue Fort Pierce, Florida 33450

Enclosure RE UEST FOR INFORMATION AUTOMATIC INITIATION OF THE AFWS AFFECT ON MAIN STEAM LINE BREAK ACCIDENT ANALYSIS A.

Return to Power 1.

Provide the results of analyses of main steam line breaks that are the most limiting with.respect to fuel failure resulting from return to power.

Analyses should be presented covering:

Break inside containment b.

Br eak outside containment c.

Availability or loss of offsite power Justify omitting an analysis for any of the above.

2. Provide the time seouence of all actions and events occurring during each of the postulated steam line break transients.

These events and actions should include:

a.

Reactor scram b.

Turbine trip c.

Steam line isolation d.

Feedwater isolation e.

ECCS actuation Auxiliary feedwater actuation and control g

Safety/relief valve actuation (prinary and secondary systems) h.

Operator actions (define credit for operator action) i.

Initiation of onsite power. (i'f required).

3; For each of the above,i identify the initiating signal, the protection system that. initiates the action, and the extent of the action ending with the time the element (i.e., l<XV, turbine stop, tur'bine control, turbine bypass, etc.)

reaches its new condition.

'Ihe above events are to reflect the

'xpected response of the plant and systems.

4.

Identify and justify any equipment that does not meet Regulatory Guides and IEEE-279 requirements.

5.

6.

Provide a list of potential sirigle failures that could affect each of the above actions and show how the analyses

'presented consider the worst single failures from a fuel failure standpoint.

Pote that normal control systems should not be considered to function if their action would be beneficial with respect to fuel failures.

Provide the following information as a function of time'.

a.

Yiinimin DNBR b.

Cladding temperature if DNBR limit is exceeded c.

Feedwater flow into faulted and nonfavlted steam generators (main and auxiliary) d.

Steam generator liquid mass, heat transfer area

covered, heat transfer rate, and pressure e

Break flow rate f.

Other steam release rates in secondary systems g.

Primary system pressure.

h.

Pressurizer level i.

Hot channel flow rate j.

Core inlet and outlet temperature k.

Pressurizer safety/relief valve flow rate 1.

ECCS flow rate.

The'analysis should be carried out until the effects of delayed neutrons and moderator feedback have turned around and the subcriticality marpin is increasing.

Note the DNBR calculations must reflect the initial plant per turhations due to'oderator and pressure decrease and'oss of offsite power (if appropriate).

Also discuss how the effects of a stuck rod are considered when calculating DNBRs after the rods have been inserted.

Xf fuel damage occurs (i.e., violation of DNBR), provide fraction of fuel that failed and offsite dose calculations.

Also provide and justify DNB correrlations used in the analyses.

B.

Containment Pressure Provide the following information to show that the containment pressure will be acceptable following a main steam line break.

1; Review your current analysis of this event, and provide NRC with the assupmtions used during this analysis.

Particular emphasis should be placed on describing how AFS flow was accounted for in your original analysis.

(Reference to previously submitted information is accep-table if identified as to page number and date.)

Any changes in your design which would impact the conclusions of your original analysis should be discussed.

We are particularly concerned with design changes that could lead'. to an underestimation of the containment pressure following a MSLB inside containment.

2.

Provide the following information for the r eanalyses performed to determine the maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels for the pro-posed AFS design.

a.

Specify the AFS flow rate that was used in your original containment pressurization analyses.

Provide the basis for this assumed flow rate.

b.

Provide the rated flow rate, the run out flow rate, and the pump head capacity curve for your AFS design.

c.

Provide the time span over which it was assumed in your original analysis that AFS was added to the affected steam generator following a

MSLB inside containment.

d.

Discuss the design provisions in the AFS used to terminate the AFS flow to the affected steam generator. If operator action is required to perform this function:; discuss the information that will be available to the operator to alert him of the need to isolate the auxiliary feedwater to the affected steam generator, the time when this information would become available, and the time it would take the operator to complete this action.

Define credit for operator action. If termination of AFS flow is dependent on automatic action, describe the basic operation of the auto-isolation system.

Describe the failure modes of the system.

Describe any annunciation devices associated with the system.

e.

Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and,energy release and the containment pressure response.

The single 'failure analysis should include, but not necess'arily be limited to:

partial loss of containment cooling systems and failure of the AFS isolation valve to close.

f.

For the single active failure case which results in the maximum containment atmosphere

pressure, provided a chronology of events.

Graphically, show the containment atmosphere pressure as a function of time for at least 30 minutes following the accident.

For this

case, assume the AFS flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current design) for the entire transient if no automatic isolation to auxiliary feedwater is part of the current design.

g.

For the case identified in (f) above, provide the mass and energy release data in tabular form.

Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.

TABLE'3;3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIOH FUNCTIOHAL UNIT EMERGENCY FEEDWATER a.

Manual b.

Steam Generator Level-Low c.

I i.i:ilestor I'low-Low d;-

Steam Generator Pressure-Low TOTAL NO.

OF CHANNELS 2 sets of 2 per FDW line 4/SG 4/FDW line 4/SG CHANNELS TO TRIP 1 setof 2

per FDW line 2/SG 2/FOW 1 ine 2/SG MINIMUM CHANNELS OPERABLE 2 sets of 2 per FDW line 3/SG 3/FDW line 3/SG APPLICABLE MODES 1,2,3,4 1, 2, 3, 4

1, 2, 3, 4

1,2,3,4 e.

Safety Injection (See Safety Injection initiating functions and requirements)

I "The provisions of Specification 3.0.4 are not applicable.

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3.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted.

This provision shall not prevent passage through OPERATIONAL MODES as required to comply with.ACTION statements.

ACTION A ACTION STATEMENTS With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or b in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ACTION B -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition

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.within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b.

All functional units receiving an input from the tripped channel are also placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4,3.2.1.

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TABLE 3.3-4 Continued ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UNIT 9.

EMLRGENCY FEEDWATER a

Mllfnlsl1 b.

Steam Generator Level-Low c.

Feedwater Flow

-Low d.

Steam Generator Pressure-Low e.

Safety In)ectlon TRIP. VALUE Not Appl)cab1e

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g Pill psia (see Safety Ingect1on Setpo)nts)

ALL'OWABLE VALUES Not Applicable 9 Pill psia

TABLE 3.3-5 Continued EhGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGN"L AND FUNCTION

RESPONSE

TIME IN SECONDS 1.

Hanual Emergency Feedwater System 2.

Steam Generator

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Pressure-Low Not Applicable Emergency Feedwater System 3.

Steam Generator

~L Emergency Feedwater System 4.

'Feedwater Flow-Low Emergency Feedwater System NOTE:

Response

time for Motor-driven Emergency Feedwater Pumps on all Safety Injectior. signal starts

• Diesel generator s arting and sequence loading delays included.

"*Diesel generator s.arting and sequence loading delays not included.

Of-,site power available.

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM'INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT l.

EMERGENCY FEEDWATER a.

Manual Initiation b.

Steam Generator Level-Low c.

I eedwater I:Iow-Low d.

Steam Generator Pressure-Low CINNNEL CHECK N.A.

CHANNEL CALIBRATION N.A.

CHANNEL I'UNCTIONAL TEST M(*)

MODES IN hNICH SURVEILLANCE RE UIRED 1, 2, 3, 4 1, 2, 3, 4 1,2,3,4 1, 2, 3, 4

e.

Safety Injection (See Safety Injection surveillance requirements)

• Manual actuation switches shall be tested at least once per 18 months during shutdown.

All other circuitry associated with manual safeguards actuation shall receive a

CHANNEL FUNCTIONAL TEST at least once per 31 days.

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