ML17301A013

PWR Main Steam Line Break W/Continued Feedwater Addition (B-69),St Lucie Unit 1, Technical Evaluation Rept

ML17301A013
Person / Time
Site:	Saint Lucie
Issue date:	11/24/1982
From:	Vosbury F FRANKLIN INSTITUTE
To:	Peter Hearn NRC
Shared Package
ML17213B015	List: ML17213B014 ML17213B016 ML17301A013
References
CON-NRC-03-81-130, CON-NRC-3-81-130 IEB-80-04, IEB-80-4, TER-C5506-138, NUDOCS 8212010094
Download: ML17301A013 (29)

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~ i Prepared by Franklin Research Center 20th and Race Street

'. Philadelphia, PA 1S103 Author:

F. V. Vosbury FRC Group Leader:

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C. Herrick Preoared for Nuclear Regulatory Commission Washington, D.C.

20555 Lead NRC Engineer:

P. Hearn November 24, 1982 aa 1

This report was prepared as an account of work sponsored by an agency of the United States Governm'ent. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information. appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

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Principal Autho Date: //-gg-S'a Reviewed by:

Group Leader Date: /l-,'n-P z Approved by Department DI ector Data:

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Franklin Research Center A Division of The Franklin Institute

'he Benjamin Franklin Parkway, Phiia PL 1 91tO (21 5) 448-1 000

E.=;CS "6 6-13 8 CONT" NTS Wh

'r'ection Title INTRODUCTlON 1.1 Purpose of Review 1.2 Generic

Background

1.3 Plant-Specific

Background

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1 3

ACCEPTANCE CRITER1A 4

ECENjCAL cMA~rUATZON 3.1 Review of Containment Pressure

Response

Analysis

3. 2 Review of React'v'y Increase Ana'ysis 3.3 Review o Corrective Actions 8

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4 CONCLUSiONS ERENCES 18 19 l7 I

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Franklin Research Center Oiv'olon ct The Frsc4din insense II

ER-C550 o-i3 8 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with'he U.S. Nuclear Regulatory Corunission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions.

The technical evaluation was conducted in accordance with criteria established by the NRC.

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W. Vosbury contributed to the technical p eparation of this report throuch a subcontract with WESTEC Services, Inc.

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RODUCTZON I

1.1 PURMSE 0" F "VZEH.

This Technical Eva3.uation Report

(~R) aocuments an independent review of Florida Power and Light Company's (FPL) compliance with the Nuclear Regulatory Comnission's (NRC)

ZE Bulletin 80-04, "Analysis of a Pzessuzized Water Reactor Main Steam Line Break with Continued Feedwater Addition" [1], as it pertains to St. Lucie Unit 1.

This evaluation was performed with the following objectives:

o to assess the conformance of FPL's main steam line break (MSLB) analyses with the reauirements of ZE Bulletin 80-04 o

to assess FPL's proposed interim and long-range corrective action plans and scheaules, if needed as a result of the MSLB analyses.

l. 2 GEIHRZC BACKGROUND Tn the summer c

1979, a pressurized water reactor (PliR) licensee su"mitted a report to the NRC that identified a cef'c'ency in the plant's o>>'i kal analys' of the containmient pressurization resulting

=zoril a MiSLB ~

A reanalys's of the containment pressure response following a MSLB was performed, and it was determined that, if the auxiliary feedwater (AFH) system continued to supply feeawater at runout conditions to the steam generator that had experienced the steam line break, containment design pressuze would be exceeded in approximately 10 minutes..

The long-term blowdown of the water supplied by the A=H system had not been considered in the earlier analysis.

On Octobei 1, 1979, the foregoing information was proviaed to all holders of ope at'ng licenses and construction permits as Z=- Znfozmation Not1ce 79-24

[2).

Another facility performed an accident analys's review pursuant to receipt of the information in the notice and discove ed that, with offsite

'electrical power available, the condensate pumps would feed the affected steam generator at an excessive rate.

This excessive feed was not previously consiaered in the plant's analysis of.a MSLB accident.

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llllll Franklin Resenrch Center ADiVialOnef The FranVI1 InSiiurte

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r 1,"I A thira licensee informed the NRC of an erzo'z in the MSLB analysis f'r their plant.

During a review of the MSLB analysis, for zero or low power at the end of core life, the licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" 'during the transient.

'Xn reality, the startup feedwater control valves will ramp to

'I 80$ full open due to an override signal resulting from the low steam genera'tor ztressure reactor trip signal.

Reanalysis of the events showed that opening of the startup valve and associated high feedwater addition to the affected*steam generator

~ould cause a rapid reactor cooldown and resultant reactor return-to-power response, a condition which 's outside the plant design basis.

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5 Review the conta'nment pressure response analys's to determine if the pocen'ti al fQL contairzilent ovezpL e ssuL e for a

. laiil s earn line break inside containment included the impact of runout tflow from the auxiliarv feecwater system and the 'mpact of other ene gy sources, such as continuation of eedwater oz condensa-e

'ow.

n vou. review, cons'er your ability to detect and isolate the camaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.

2 ~

Review your analysis of the reactivity increase which rt:suits from a at ain steam line break inside or outside containment.

This review'hould consider the reactor cooldown zate and,the potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position. lf your previous analysis did not consider

.all potential water sources,(such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated the report of this review should include:

a.

The boundary conditions for the analysis, e.g.,

the end of 1'fe shutdown marg'n, the moderator temperature coefficient, power

, level and the net effect of the associate'd steam generator water inventory on the reactor system cooling, etc.,

Because of these deficiencies identified in original MSLB accident

'analyses, the HR issued ZE Bulletin 80-04 on February 8,

1980-This bulletin

', required all PRRs with operating licenses and certain near-term PHR operating license applicants to perform the following:

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The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the celivery of high concentration boric acid solution to the reactor coolant system,

T:-~-C550 6-a 8 c.

he effect o extenced wate supp'y "o the af ec"ea steam genera or on the core crit'cality ana etu n

o power, N 5

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N d.

The hot channel factors corresponding to the most reactive rod in the fully witharawn position at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient.

3. If the potential for containment overpressure exists or the reactor return-to-power response

worsens, provide a proposed corrective action and a schedule for completion of the corrective action.

If'he unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."

l. 3 Pr~hZ-SPECI" IC BACKGROUND

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-PL responaea to IE Bulletin 80-04 in a letter to the NRC aated May 8, 1980

[3) and provided additional information in letters dated July 23, 1981

[4) ana October 20, 1981 [5).

On August 9, 1982 [6), the Licensee provided acdit'onal recuested informat'on recuired to complete this review.

The

'.". ormat'on nn References 3 f 4/ 5, pe t'nent information from the St.

and 6 has been evaluated along with Lucie Unit 1 upaatea

"-ideal Safety Analysis

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[7) "o determine he adecuacy of the Licensee's compliance with B"'let'n 80-04.

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2.

ACCEPTANCE CR:TERTA The follow'ng criteria against which the Licensee's MSLB response was evaluated were provided'by the NRC f8):

1.

PHR licensees'esponses to IE Bulletin 80-04 shall include. the following information related to their analysis 'of containment pressure and core reactivity response to a MSLB within or outside con ta inmen t:

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a.

A discussion of the continuation of flow to the affected steam,,

generator, including the impact of runout flow from the APW system and the impact of other energy sources, such as continuation of feedwate" or concensate flow.

A=el system runout flow should be determined from the manufacturer's pump curves at no backpressure, unless the system contains reliable anti-runout provisions or a more representative backpressure has been conservatively calculated.

Ef a licensee assumes credit for

'anti-runout provisions, then justification and/or documentation-.

used to determine that the provisions are reliable should be provided.

Examples of devices for wnich provisions are= reliable a

e ant'- unout devices that use act"e components (e.c.,

autor:.atically throttled valves) which meet t'.".e. require. ents of IEEE Std 279-1971 (9] and passive devices (e.a.,

low orifi'ces or cavitating venturis)

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b.

A cete mination of potent'al containment overpressure as a result of the impact of runout flow from De A'M system or De impact of other energy sources such as continuation of feedwater or condensate flow.

Where a revised analysis is submitted or where reference is made to the existing PSAR analysis, the-analysis must show that runout A'rvl flow was included and that design containment pressure was not exceeded.

c.

A discussion of the ability to detect and isolate the damaged steam generator from continued feedwater addition during the MSLB acc'dent.

Operator action to isolate AFH flow to the affected steam generator within the first 30 minutes of the sta t of the MSLB should be justified.

Zf operator act'on is to be completed within the first 10 minutes, then the justification should address the indication available to the operator and the actions

.required.

Where operator action is reduired to prevent exceeding a design value, i.e., containment design pressure or specified acceptable fuel design 1'mits, then the discussion should include the calculated time when the design value, would be exceeded if no operator action were assumed.

Where operator actions are to be performed between 10 and 30 minutes after the start of the MSLB, the justification should address the indications available 'to the operator and the operator actions required, noting that for the Ill)(l FranMin Research Center A Dinvon or The Franktin lrentutc

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ac 'o..s

~ oula be pe formed rom the control room.

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Nhere all water sources were not considered in the previous analysis, an indication should be provided of the core reactivity change which'esults from the inclusion of aaditional water sources.

A submittal which does not aetermine the magnitude of reactivity change from an original analysis is not responsive to the requirements of IE Bulletin 80-04.

2 ~ If containment overpressure or a worsening of the reactor return-to-power wiA a violation of the specified acceptable fuel design limits described in Section 4.2 of the Standard Review Plan

[10]

(i.e., increase in core reactivity) can occur by the licensee's

analysis, the licensee shall provide the following additional information:

a.

The proposed corrective actions to prevent containment overpressure or the violation of fuel design limits, and the schedule for their completion.

b.

The interim actions that will be taken until the proposea corrective action is co.pleted, if the un't 's operating.

The acceptable input assumptions used in

~~e 1'censee's analysis o

the core reactivity changes during a

NSLB are given in Section 15;1.5.

of the Standara Review Plan [11].

The fo3.lowing spec'c assump ions should be used unless the ana'ysis shows that a ai ferent assunption i.s more limiting:

Assumption II.3.b.:

Analysis should be performed to dete mine the most conservative assumption with respe'ct to a loss of elec""ical power.

A reactivity analysis shoula be conducted for a normal power situation as well as a loss of.offsite power scenar'o, unless the licensee has previously conducted a sensitivity analysis which demonstrates that a particular assumption i.s nore conservative.

Assumption Il.3.d.:

The most restrictive single active failure in the safety injection system which has the effect of delaying /he delivery of high concentration boric acid solut'on to the reactor coolant system, or any other single

'ctive failure affecting the plant response, shoula be considered.

Assumption II.3.g.:

The in'tial core flow should be chosen such that the post&SLB shutdown margin is minimized (i.e.,

maximum initial core flow).

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The acceptable computer coces -or the lhcensee's analysis o

core reactivity changes

are, by nuclear steam supply system (NSSS)

venaor, the folTowing:

CESEC (Combustion Enaineering),

MrPTBAN (Westing-house),

ana TRAP (Babcock G Wilcox)..Other computer" codes may be

used, provided that these codes have previously been reviewed and found to be acceptable by the NBC staff. If a computer code is used which has not been reviewed, the licensee must describe the method employed to Verify the code results in sufficient detail to permit the coae to be reviewed for acceptability.

If the APW. pumps can be damaged by extended operation at runout flow, the licensee's action to preclude damage should be reviewed for technical merit.

Any active features should satisfy the requirements of IEEE Std 279-1971.

Where no corrective action has been proposed, this should be indicated to the NRC for further action and resolution.

Modifications to electrical instrumentation and controls needed to detect and initiate isolation of the affected steam generator ana feedwater sources in order to prevent containment overpressure and/or unacceptable core reactivity increases must satisfy safety-grade recu'remen s.

Instrumentation that the operator rel'es upon to follow the accident and to aetermine isolation of the affectea st am cenerato and eedwater sources should conform to the criteria contained in ANS/ANSI-4.5-1980, "Criter'a o" Accident Moiitor'ng

=-unctions in Light-Water-Cooled Reactors"

[12), and the regulato y positions in Regulatory Guide 1.97, Bev.

2a

~ nstrumentation for Light-Water-Cooled Nuc'ear Power Plants to Assess Plant and Env'rons Cond't'ons During and =<<ollowing an Accident" [13).

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A&A system status should be reviewed to ensure that system heat removal capacity does not decrease below the minimum reauired level as a result of isolation of the affected steam generator; and also t'1Tat recent changes have not been made in the system which adverse'y affect vital assumptions of the containment pressure and core reactivity response analyses.

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.The safety-arade requirements (redundancy, seismic ana environmental qualificat'ons, etc.) of the eauipment that isolates the main feedwater (K8) and A:-W systems from the affected steam generator should be specified.

The moaifications of equipment that is relied upon to isolate the FEW and AiW systems from the affected steam generator should satisfy the following criteria to be considered safety-grade:

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Redundancy and power source requirements:

The isolation valves should be aesig'ned to accommodate a single failure.

A failure-modes-and-effects analysis should demonstrate that the system is capable of withstanding a single failure without loss of function.

The single failure analysis should be conducted in accordance

TE<<-C550=--'". 8 with the appropria"e ru'es of app'cat'on o

ANS-51.7/H658-1976, "Single:ailure Criteria for FnR "-'uid Systems"

[14).

o Seismic recuirements:

The isolation valves should be designed to Cat'egory T. as recommended in Regulatory Guide.1.26

$15).

o Environmental qualification:

.The isolation valves should 'satisfy the requirements of NUR=-G-0588, Rev.

1, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment"

[16].

o Quality standards:

The isolation valves should satisfy Group 3 quality standards as recommended in Regulatory Guide 1.26 or similar quality standards from the p3,ant's licensing bases.

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Division ot The Fnn)hn Insotutc

TER-C5506-138 3.

T" C-lICAL "VALUATION I

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4 The scope of work included the ollowing:

1.

Review the Licensee's response o IE Bulletin 80-04 against the acceptance criteria.

2.

a.

Evaluate the Licensee's HSLB analyses for the potential of overpressurizing the containment and with respect to the core reactivity increase due to the effect of continued feedwater flow.

b.

Evaluate the Licensee's proposed corrective actions and scheaule for implementation if the findings of Task 2a indicate that a' potential exists for cverpressurizing the containment or worsening the reactor return-to-power in the event of a KALB acc'dent.

3.

Prepare a TER for each plant based on the evaluation of the in ormation presented for Tasks 1 and 2 above.

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1 Th's report constitutes a TER in sat's action of Task 3.

Sect'ons 3.1 trough 3.3 of this reoort state t'.".e recuirements o

"= BI llet'n 80-04 bv suosection, summa ize the Licensee's statements ana conclusions recara'nc

"".ese requ'rements, anc present a d'sc ssion of the

~ icensee's evaluation followed by conclusions ana recommendations.

3.1 REVI "vf OF CONTAI&KViZPRESSUR=

RESPONSE

ANALYSIS r

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~ 4 The requirement from IE Bullet'n 80-04, Item 1, is as follows:

~ "Rev'ew the containment pressure response analysis to aetermine if the ootential for containment overpressure for a main steam line break inside containment incluaed the impact of runout flow from the auxiliary feeawater system and the impac" of other energy sources, such as continuation of feedwater or condensate flow.

'In your review, consider your ability to detect and isolate Ne damaged steam generator f om these sources and the ability of the pumps to remain ooerable after extended operat'on at runout flow."

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3.1'.1 Summary of Licensee Statemen" s and Conclusions I

'In regard to the review of the containment pressure response

analysis, the Licensee stated

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"'ie have rev'ewed tne containment respcnse analvsis for a ma'n steam line break (as containea in the St.

~ ucie Unit 1

"-SAR at Subsection 6.2.1.3.2 (c)...),

ana we have aetermined that "n~ potential for containment pressurizat'on c'd in act aadress the ir pact of add'ional.flow from the auxiliary or main feedwater system.

Pertinent discussions presented in

'he PSAR are reiterated herein for your information.

The main feedwater isolation valves (E-MV-09-7, I-MV-09-8) are aesigned to close in 60 seconas or less and are surveillance-tested in accordance with the plant Technical Specifications.

Moreover both the main feedwater isolation valves and the main feedwater pump isolation valves,'HV-09-1, MV-09-2) close on receipt of a Safety Injection Actuation Signal (SEAS) and/or a Main Steam Xsolation Signal (MSIS).

Thus for a main s earn line break inside containment (the. concern of IE Bulletin 80-04), main feedwater flow will be terminated very early in the transient br an SEAS which is generated at abou" 2 seconas.

he present Auxil'ary <<eeawater System (P=-RS) for St. Lucie Unit 1 is manual'y initiated within 13 minutes; since the ooerator has ample time to distinguish the faulted steam generator by observing the redundant Class

= steam generator pressure and level inaicat'ons, there will be no Auxiliarv =eeawater adaea. to the faulted stean generator.

Analyses are underway to dete m'ne the effects of NP -mancated installation of autor:.atically in'tated P.8 low.

Xf the rest~Its indica e contain.-..ent over-pressurization, retention or the manual A:-8 flow init'at'on shou'a be a cons'aeration.

The following discussons are extractea directly f om "'".e St.

Lucie Un'" 1 PS& at paces (6.2-27 and 6.2-28].

The closure time for the steam line isolation valves is assumed to be a

six second ramp that initiates on MSIS and the feedwater valves close on a sixty second ramp (testing has shown the feedwater isolation valves close in 20 seconds so that 60 second ramp is conservative) initiated by MSXS'or SIAS.

The code assumes closure of all steam generator isolation valves on MSIS for the faulted steam generator.

In addition, back flow into the containment from the unfaulted steam generator via the main steamline header crosstie is assumed until isolation valve closure occurs.

Assuming the worst case, i.e. single failure of the feeawater block valve (I-MV-09-7 or -8), the pump isolation valve closes in 60 seconds after receipt of MSXS or SEAS and approximately 107,000 pounas of feedwater are adaed to the faulted steam generator.

The analysis conservatively assumed that the main eed pump operates at runout flow during valve closure and fur Her that all of the high temperature fluid in the feecwater lines to the furthest isolation valve flashes into the containment.

The highest containment pressure.

and temperature i's found for the case of the 105 percent power (2698 MNT), 85 percent break area (5.355 ft2) steam line break."

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Subsecuent to the L'censee's response to "-

ulle in 80-04'

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"Calculations have been performed to determine if the containment building response (pressure) and the core reactivity response (return to power) are acceptable following a main steam line break inside containment when auxiliary feedwater is added without regard to the identification of the. affected steam generator.

Both the containment pressure and the reactivity values were found to be within acceptable 1imits.'"

svstem was upgraded to be automatically initiated.

Xn recara to the validity of the anal sis Section 10.5.3 of the u dated PS~% states 7

3;1.2

. valuation he Licensee's submittals

[3-6] concerning the containment pressure response fol'owing a HSM and applicable sections of the St. Lucie Plant

""SAR.

[7) were reviewed in order to evaluate whether the follow'ng portions of the acceptance criteria were met:

O'riterion l.a Continuat'on of flow to the, affected steam generator o

Criterion l.b Potential for containment overpressure o

C iter ion '

c Ab'1'y to detect and 'olate

=he damagec s" earn generator o

Criterion 4

o Criterion 5

Potential for AFW pump damage Design of steam and feedwater isolation system o

Criter'on 6

o Criterion 7

Decay heat removal capacity II Safety-grade requirements for NH< and A.=v7 isolation valves.

St. Lucie Unit 1 is a Combustion Engineering-designed, 2-loop, 2700-Nht plant.

~;he following svstems provide the necessary protection against a steam pipe rupture:

o Safety injection actuation signal (SIAS) generated on:

a.

two out of four (2/4) low pressurizer pressure signals (1600 psig) b.

2/4 high containment pressure signals (5 psig) U UUI! Franklin Research Center A CHvision d The Frank'nsctute

TE"";C550 6-13 8 o

Reacto'." tzip on:

a.

high neutron lux (107%)

b.

high containment pressure (3.3 psig) c.

low steam generator pressure (585 psig) o Main steam isolation signal (MSZS), generated on 2/4 low steam generator pressure signals (485 psig), trips safety-grade, fast-acting main steam isolation valves (designed to close in 6 seconds) o Redundant isolation of MPR lines.

Both the MFA block valves (safety-grade) and M&I pump isolation valves (designed to close in 60, seconds) close on receipt of a SZAS and/or gSZS 1

o Containment spray is actuated on high containr:.ent pressure (10 psig) coincident with a SZAS.

Each steam line has a fast-closing stop valve with downstream check valve.

These four valves prevent blowdown of more than one steara generator for any break location even if one valve fails to close.

Foz breaks upstream of the stop valve 'n one 1'ne, closure of either the check valve 'n that line or the stop valve 'n the othe line will prevent blowdown of the other stea generator.

For all breaks, th's arrangement precludes b'owcown oz more han one steam generator inside tne containment.

Howevez, the analysis coes not take credit for the check valves and assumes reverse flow from the intact steam generator to the rupture until the MSZS causes both steam isolation valves to shut.

The AFH system consists of a single turbine-driven pump (500 gpm) and two pump suppl'ies both steam generators and each motor-driven pump supplies one steam generator.

The AF8 flow from any one pump will ensure that the heat removal capacity exceeds the minimum level required for decav heat removal aftez a MSLB.

1 After the Licensee's initial res onse to ZE Bulleti 4.

P n 80 04 f3],

he AEW systera was upgraaed from the original manually initiated version.

The new auxiliary feegwater actuation. system (AFAS) satisfies the short-term rec(uirements of NUR~~-0737 [17), paragraph ZZ ~E.l 2g in that the AFAS is a motor-,driven pumps (350 gpm each)

< and is aligned so that the turbine-driven

-ll-Franklin Research Center Drvisen d Thc Franl4n Insiaae

TER-C5506-U8 contro'-grade svstem.

The A"-AS vill automaticallv initiate both the motor-criven and turb'ne-driven pumps after a 3-minute delay upon receipt of 2/4 low steam generator level signals for either steam generator."

The engineered safety features actuation system (ESEAS), which includes SIAS and MSIS systems,.

meets safety-grade and IZEE Std 279-1971 reauirements.

The environmental clualification of safety-related electrical and mechanical components is being reviewed separately by the NRC and is not within the scope of this review.

The ESAR analysis determined that the worst-case HSLB vas that which 2

assumed an 85% break area (5.355 ft ), 105si full power, and a sine'le failure of a feecvater block valve, which allowed 107,000 pounds of eedwa er to be

.added to the affected steam generator until the bEH pump isolation valve closed at 60 seconds.

The analysis assumed that the

~i% pump ope ates at runout flow until the isolat:ion valve is fully closed.

This MSLB analysis oroducec a -eak =ressure o

41.6 =sig at 120 seconds.

Tl..e initiation at 180 seconds does not a feet the peak pressu e produced.

The MSL3 analvsis assumed that the operator 'solates A:-R flow to the ruptured steam generator at 10 minutes after the start of the MSLH.

The operator has sufficient instrumentation to analyze the accident and determine which steam generator should be isolated.

The operator's actions are minimal and can be performed in the control room within the 10 minutes.

The review did not determine if the instrumentation upon which the operator relies to follow the accident and isolate the affected steam generator conforms with the criteria in ANS/MSI-4.5-1980

[12) and Regulatory Guice 1.97 f13) ~

In order to comply vith the long-term recuirements of Item II.-.1.2 of NUR=-G-0737, the Licensee intends to install an APM that meets safety-grade and IEEE Std 279-1971 reau

ements.

This safety-grade ALAS will initiate AEW to the intact steam generator immediately upon demand and terminate or prevent AW flow to the steam generator identified as being ruptured.

A ruptured steam generator is identified by the following conditions:

llÃ'ranklinResearch Center A Division or Thc FmnkfinI~

T:-R-C550 6-'3 8 o

one steam cenerator pressure

's 100 ps'c below the othe steam genera or p essure or o

one feedwater supply header pressure is 100 psig below the other feedwatez supply header pressure and o

water level is low in the steam generator with the lower pressure and o

the other steam generator oz feedwater header is not identified as being ruptured.

Since the sa ety-grade ~AS is designed to direct flow only to the unaffected steam generator, the assumptions of the FSAR HSLB analysis will emain conservative, and the ABf pumps will be protected from runout flow.

The present control-grade APAS system does not protect the AEH pumps from the e==ects of runou

=low.

The sa etv-crade A"-M 's to be installed duzinc the r'e ueling outage scheduled to start in March 1983 [6].

3.1.3 Conclusion The Licensee's responses

[3-6] and the St. Lucie PSAR [7] adequately address the concerns of Item 1 of IE Bulletin 80-04.

The containment pressure response analysis and the design of the mitigating systems satisfy the NRC's acceptance criteria.

Regarding item 1, it is concluded that there is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition.

The installation of the safety-grade AFAS will

'ncrease the margin between containment design pressure and the maximum pressure obtained during the transient and will eliminate the need for operator ac-ion to isolate auxiliary feedwater to the ruptured steam generator.

The control-grade AFAS does not protect tne APH pumps from runout flow.

The installation of the safety-grade A=AS will provide the AB7 pumps with adequate runout protection.

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~ '"1 The requirement from ZE Bulletin 80-04, Ztem 2, is as follows:

I "Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment.

This review should consider the reactor cooldown rate and the potential for the reactor to return-to-power with the most reactive control rod 'in the fully withdrawn position.

Zf your previous analysis did not consider all potential wat'er sources (such as those listed in 1 above) and if the reactivity increase i.s greater than previous analysis indicated the report of this review should include:

'a 0 The boundary conditions for the analysis, e.g.,

the end of life shutaown margin, the moderator temperature coefficient, power 'evel and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,

11 1

b.

The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the aelive y of hicn concentra"ion boric acid solution to the reactor coolant system, c ~

The e

ect of extended water supply to the affected steam generator or. the co e c iticality ana retuzn--c-powe

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3.2.1 Summary of Zicensee Statements and Conclusionh he hot channel factozs corresponding to the mos react've roa in the fully withdrawn position at the ena. of life, ana the Minimum Doparture from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient."

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i richen the control-grade APAS was installed, the Licensee performed an analysis to aetermine the effects oz APH initiation on the reactivity increase after a MSLB.. This analysis was initially submitted in Refe ence 4 and late incorporated into Section 15.4.6 of the updated FSAR [7).

The analysis considered two cases:

o two-loop full power, 2754 MWt o

two-loop no load, 1 MWt.

The results o

the full'ower analvsis show that a

SZAS is actuated at 15

seconds, at which time the reactor coolant pumps (R:Ps) are manually tripped by the operator.

The manual trip of RCPs slows down the rate of primary heat removal and thus delays the time when the affected steam generator blows dzy..

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earn generator blows czy at 154 seconas ana terminates the cool-,

aown of the R~S.

The peak eactivity attained p"ior to delivery of A=-W flow is -0.205% at 169 seconds.

A peak post-trip powez of 14.0%, consisting of 12,2e decay heat and 1.8% fission power, is produced at 72.8 seconds.

The continued production of decay heat from the fuel after termination of blowdown causes the reactor coolant temperatures to increase.

This in turn reduces the magnitude of the positive moderator reactivity inserted, and thus the total.

reactivity becomes hore negative.

The delivery of AZW flow to the affected steam generator at 180 seconds

, initiates a further cooldown of the RCS, which results in more positive reactivity insertion.

However, continued addition of negative reactivity by the boron injected via the high-pressure safety injection pump prevents the positive re'activity insertion of the moderator from causing a z'etuzn to criticali"y.

The core remains subcritical throughout the remainaer of the transient.

The steam line rupture event f om hot full power conait'ons w'th automatic initiation of A=A and manual

" ip of KPs on a

SZAS shcws that

".'he coze coes not pioauce sign'cant ins antaneous fission power.

Since

"-here is no significant return-to-power, it'an be concluded that critical heat fluxes will not be exceeded.

The fesults of the no-load analysis show that a SIAS is actuated at 12.0 seconds-,

at which time the R:Ps are manually tripped by the operator.

The affected steam generator blows dry at 180.0 seconds.

ABC flow is initiated at

'I 180 1 seconds, which continues the cooldown of the RCS.

Thus, tHe total core reactivity is critical for a short period of time.

The peak reactivity atta'ned is only +0.140t at 401.0 seconds.

The addition of boron via high-pressure safety injection mitigates the reactivity transient.

The core power 1

level remain5 less than 18 of rated power at all times during the event.

Since there is no significant return-tc power for the two loopfno load case, it can be concluded that the critical heat flux will not be exceeded during this event.

llllrl Franklin Research center A Division ol %bc FesnkSn Insane

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-value ion The Licensee's PS~K analysis of the core reactivity increase resulting from a MSLB with contiqued feedwater addition was reviewed in order to evaluate whether the fo3.1owi'ng acceptance criteria were met:

o Criterion l.c' Ability to detect and isolate the damaged steam generator o

Criterion l.d -

Changes in core reactivity increase o'riterion 3

Analysis assumptions.

Prom that review, it was determined that the analysis is conservative in its

,-assumptions and that the assumptions a

e in accordance with those in Acceptance Criterion 3.

The two-loop full power MSLB produced the most severe return-to-power

, t ansient wi h a peak power of 14$

(12.2e decay heat and 1.8% fission power occur'ng at 72.8 seconcs.

The minimum subcritical margin obtained was

-0.205't 169 secor.ds.

The celivery of APt~ at 180 seconds results in the insertion o

"cs't've eac" ivity, but the cont nued addition of bo on 0

he core p even" s a'et rn-to-power.

The return-to-power transient does not violate the specified acceptable fuel design limits.

The installation of the safety-grade APAS will require a reanalysis of the reactivity transient because the earlier addi.tion of AFH to the unaffected steam generator will cause a greater cooldown rate in'tially.

The Licensee committed

[6) to complete thi.s analysis in December

1982, which is prior to the*scheduled installation of the sa ety-grade APAS.

3.2.3 Conclusion cj The Licensee's response and PS'deauately address the concerns of Item 2 of 7-Bulletin 80-04.

All potential sources of water were identified, and althougn a reactor return-tc power is predicted, there is no violation of the

.~ specif'ed acceptable fuel design limits.

Therefore, the PSALM analysis of the i

.-, reacti'vity increase resulting from a MSLB remains valid.

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llB FranMin Research Center ADivan d The Fatkhn Inscnutc

=R-C5 v0 6-~ 8 3.3 5"VI=4 0" COB.'ACTI%- ACTIONS The recuirement from IE Bulletin 80-04+ Item 3, is as follows:

"If the potential for containment overpressure exists or the reactor return-to-power response

worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."

'3.3.1 Summary of L'icensee Statements and Conclusions In regard to the need for a reanalysis of the reactivity response to a MSLB with the safety-grade A=AS ins alled, the Licensee stated

[6]:

"He have contracted with our new fuel vendor to perform a MSLB analysis using the parameters of the 'sraart'PH system as input.

This analysis is due to be completed in December 1982, at which time we will provide a

response to your concern.

This schedule is justified by the fact that the system is not scheduled

".o be completely installed until our next ef el'ng ou ac wh'ch is

==resently schecu'ed or March 1983."

3.3.2.

- Jaluation and Conclusion The Licensee's analysis dete min'ed that, for the current ArM system design, neither a containment overpressurization nor a reactor return-to-power with a resultant violation of. the specified acceptable fuel design limits would occur from a MSLB.

The Licensee's commitment to provide an analysis of the reactivity response prior to the installation of the safety-grade UPAS is acceptable.

The installation of the safety-grade A~AS will also provide runout protection for the AFR pumps. ~

Ul!Il Franklin Research Center h Dins'en d ice FruJdin Insetutc

=R-C5=06-l38 4.

COYCLUSTOssS Conclusions regarding Florida Power and Light Company's response to EE Bulletin 80-04 relative'to St. Lucie Unit l are as follows:

o There is no potential for containment overpressurization resulting from a main steam line break (MSLB) with continued feedwater addition.

o All potential watch sources were identified and, although a reactor return-to-power is predicted, there is no violation of the specified.

acceptable fuel design limits.

Therefore, the Updated rinal Safety Analysis Report MSLB reactivity increase analysis remains valid for the control-grade auxiliary feedwater actuation system (AZAS).

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The cont ol-grade APAS does not provide runout protection for the auxiliary feecwater pumps.

Enstallation of the safety-grace APAS will provide unout protection.

o The Licensee's commitment to provide an analysis of the reactivity response to a MSLB prior to installation of the sa ety-grade A=AS is acceptable.

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