L-79-287, Responds to NRC & IE Info Notice 79-22.Review of Postulated Interactions Indicates No Substantial Safety Hazard

Responds to NRC & IE Info Notice 79-22.Review of Postulated Interactions Indicates No Substantial Safety Hazard

ML17207A462
Person / Time
Site:	Saint Lucie
Issue date:	10/08/1979
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
L-79-287, NUDOCS 7910120260
Download: ML17207A462 (43)

Text

TOTAL NUMBER OF COVIES REQUIRED:

LTTR 38 ENCL 37 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR-'7910120260 DOC ~ DATE: 79/10/08 NOTARIZED:

NO FACIL:50-335 St ~ Lucie Plant<

Unit ii Flor ida Power L Light Co.

AUTH,NAME AUTHOR AFF I'LIATION UHRIGg R,G ~

Flor ida Power 8 Light Co, REC IP ~ NAME RECIPIENT AFFILIATION DENTONrH ~ RE Office of Nuclear Reactor Regulation

SUBJECT:

'esponds to NRC 790910 ltr re interactions between nonsafety L safety grade equipment, util 8 vendor reviewed postulated interactions,No substantial safety hazard.

DOCKET 05000335 DISTRIBUTION CODE:

A038S COPIES RECEIVED:LTR ENCL +

SIZE:

TITLE: Resp 9/17/79 Denton Ltr-Interact Sf ty Gr de Sys 8

Non Sgs OTES W

~ >>>>>>>>e>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>were>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>>

RECIPIENT ID CODE/NAME ACTION; 5

BC ORB WV INTERNA 11 REAC SFTY BR 10 EEB 16 OELD 3

LPDR 6

T MARSH 8

H MORRIS EXTERNAL:

17 ACRS COPIES LTTR ENCL 1

1 1

1 1

1 1

1 1

1 1

1 16 16 RECIPIENT ID CODF/NAME 10 ENGR BR 12 ILE 15 EFLT TRT SYS 2

NRC PDR 5

M GROTENHUIS 7 J ROSENTHAL 9

CORE PERF ER

~rs vsNk NSIC COPIES LTTR ENCL 1

1 2

2 1

1 1

1 1

1 1

1 1

1 r

I 1

1 OCT 16 1979

0 n

'I d

I t

AP pl

4'

. BOX 526100, MIAMI,FL 33162

~q)ll<q~

FLORIDA POWER 8 LIGHTCOMPANY October 8, 1979 L-79-287 Nr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.

S. Nuclear Regulatory Commission Washington, D.

C.

20555

Dear Nr. Denton:

Re:

St. Lucie Unit 1

Docket No. 50-335, Safet /Control Interactions This letter is submitted in response to your letter of September 17, 1979 on the subject of postulated interactions between non-safety grade systems and safety grade systems.

This subject was also addressed in IE Information Notice 79-22 issued on September 14, 1979.

Florida Power 5 Light Company, in conjunction with Combustion Engineering, has reviewed both the specific systems listed in IE Information Notice 79-22 and other non-safety grade systems which could possibly have the potential to interact with safety grade systems.

In no instance have we been able to identify any such potential interaction which could constitute a substantial safety hazard.

The information contained in this letter responds to the concerns of the NRC on this subject and justifies continued operation of St. Luci e Unit 1 on the basis of the inapplicability and/or the improbability of the postulated scenario relative to the St. Luci e design.

Analyses recently performed by NSAC and others conclude that the probability of a high energy line break at a typical nuclear plant is very low.

Furthermore, any such break is far, more likely to be a small crack rather than a major line break.

Additionally, our review recognized the difference between a licensing deficiency and a substantial safety hazard.

Despite the low probability of a high energy line break, Combustion Engineering has performed a review of thirteen systems involving five accident scenarios which encompass the spectrum of postulated high energy line breaks.

This evaluation is attached as Appendix A.

Of the total of sixty-five possible combinations of control systems and possible accidents, fourteen scenarios (identified by an X on Table 1) were identified as warranting further review.

These fourteen generic concerns are bounded by the seven plant-specific interactions discussed in Appendix B, which is an evaluation performed by FPL particular to the design of St. Lucie Unit 1.

The conclusions reached therein support and justify the continued operation of the unit.

FPL will attempt to resolve any significant items identified in this or future submittals within the following schedule:

791912 ~

PEOPLE... SERVING PEOPLE

A f,c Ji

Mr. Harold R. Denton, Director Page 2

items whose solutions do not require equipment modifications (i.e.,

procedural

changes, training, etc.)

are scheduled for resolution by January 1, 1980.

items whose solutions requi re equipment modifications are scheduled for resolution by June 1, 1981.

Very truly o rs, Robert E.

hrig Vice President Advanced Systems 8 Technology REU/MAS/DKJ/cph Appendices (2) cc:

Mr. James P. O'Reilly, Region II Robert Lowenstein, Esquire

0 i

i b

u+

'1

'C i

~

~

~

~

~

I

~

~

~

~

I ~"

~ I

~

I I

I I

I

~

~

~

I ~

I I

~

~

~

~

~

~ ~

I

~

~

~

~

p 4

APPENDIX A Re:

St. Lucie Unit l Docket No. 50-335 Safet /Control Interactions AR F

INTERACTION SCENARIOS

At the request of the C-E Owners Group on Post-THI Efforts, Combustion

'ngineering conducted a review of potential control systems interactions during high energy pipe break events.

C-E initially established a matrix (Table 1) of high energy pipe break events and control functions within C-E's ability to properly evaluate.

A list of separate systems that should also be con-sidered was developed and forwarded to the utilities participating in this effort along with a list of control function and events under consideration by C-E.

These lists are attached.

In the, time 'available, C-E reduced this-.-*matri.x to include only. those

.systems and events-which 'required'urther evaluation.

'Some of'hese events may be eliminated by individual. util,ities on a plant specific basis.

A gen-

eral description of the procedure used'by C-E'--to -reduce thfs matrix is-listed below.

I.

An initial review of each postulated Control Function failure.-

for each pipe break was compTeted and.served:

as the basis for consideration.

1lhere a postulated fai'lure-couTd potentially increase the severity of'a high energy pipe break, the following criteria were employed to resolve the concern:,

1.

Is the postulated Control Function failure mode credible' 2.

Is the Control Function Equipment (Sensor, Cables, etc:)

in a location which could be impacted by the environment?

3.

Is the Control Function Equipment (Sensor, Cable, etc.)

qualified to operate properly in the postulated environment?

4.

Hhere the postulated Control Function failure is credible, could its impact potentially affect the conclusions pre-sented in the SAR?

Considerations such as Haximum Control Function capabilities, and delayed, but proper operator action were employed in this effort; Prudent engineering judgement was utilized to eliminate those events/

interactions which did not change. the conclusions of SAR analyses.'xtensive evaluations

.invoTm'ng: the. Auxiliary Eeedwater.

system and other long term cooling mechanisms have; not--been-.performed;"

Auxili'ary feedwater ss=-

being evaluated under Bulletins and Orders.and Lessons Learned'(hUREG 0578)

Task Forces.=.-This decision

!~as: made in order to concentrate'n those items felt to be of greater sionificance for assessment of control system hioh energy pipe break interactions.

In several

cases, most notibly the PORV failure in the open position, no specific failure mechanism has been identified.

The only manner for such

a failure to occur would be for power to be inadvertently applied to the valv'e solenoid and not be removed.

Part of C-E's short term recommendations are for utilities to evaluate whether or not a failure mechanism of this type is credible.

The potential adverse impact of high energy pipe breaks on reactor coolant pumps was considered.

Both the seized shaft and the simultaneous three or four pump loss of flow were eliminated from consideration based on judge-ment that these failures-are not considered. credible within the time frame limited by operator action (30. minutes}'ue to envfronmental impact alone.

The impact of'ther potential loss of. flow events (e.g.

one or two pump loss of

'low) during high energy pipe breaks was reviewed and it was judged that the resulting rapid reactor trip was sufficient to ensure that the conclusions of the SAR would not change.

Attachments detail specific event/interaction scenarios and define specific short term recommendations which have been established to minimize the probability and impact of the postuTated events.

These Attachments also discuss potential long term alternatives which have been identified.

I

~

r I

CONTROL FUNCTIONS AND EVENTS Control Functions Considered Pressurizer Level Pressurizer Pressure PORV's 8 Block Valves RCP's Rod Position (e.g.

RRS,.CEDS):-f.Pretrip]

Boron Concentration (CVCS)

,Feedwate'r Flow (FHCS)

Steam Flow to Turbine (Turbine C.S.)

Steam By-Pass to Condenser (SBCS)

Steam Dumps.to Atmosphere Downstream of HSIV's Steam Dumps to Atmosphere Upstream of'f<SIV's

. Steam Generator Blowdown

-,Assumption of Loss of.Condenser-Because of

- Condensate Transfer

- Condenser Operation The listed systems will be evaluated in conjunction with the following events:

Small Steamline Rupture Inside Containm nt Small Steamline Rupture Outside Containment Large Steamline Rupture Inside Containment Large Steamline Rupture Outside Containment Small Feedline Rupture. Inside Containment Small Feedline,Rupture Outside:-Containment Large Feedline Rupture-I'nsfde-Contaa"nment Large Feedline Rupture-Outs>de..Co'ntainm nt.,:

Small LOCA Inside Containment Small LOCA Outside Containment Large LOCA Pod Ejection

, DESCRIPTIO RENAINING EVENTS AND CONTROL iNCTIONS I.

Assessment of Control System Failures on Steam Line Break Event A.

~Sequence of Events for Generic SAR'teara Line Break at Full Pouer, Inside or Outside Containment, 1.

Double-ended steam line break occurs' 2.

Reactor trip on low stean generator pressure 3.

tlSIS initiates to isolate the steam generators 4.

RCS temperature decreases due -to excessive steam removal

. 5-.

Total reactivity increases; due:.'to--moderator cooldown effect -..'-,

~-

6.

tlSIVs close a

7.

Pressurizer empties 8.

Low pressurizer pressure initiates SIAS 9.

HFIVs close

= 10.

Safety injection boron 'reaches.core ll.

Affected steam generator

empties, terminating cooldown effect,the transient reactivity reaches peak and decreases gradually due to boron ingection 12-Limited or no post-trip return-to-power 13; No fuel in DNB B.

Steam Line Break With PORV Control S stem Failure 1.

Significant Interaction Effects:

a.

Increased Containment Pressure b.

-A stuck open PORV in combination with a steam line break has not 1 een analyzed.

2.

Assumptions a.

Steam line break (large-brea'k inside..containment, for any size or location for. Item 1.8'bove},.

E b.

Inadvertently'ORVs..open.'and; iemaTn. open.

c.

PORV Block valve also;faHs-to close

ihen required.

d.

Initial condition:.full power Item-1.A above,,

3.

It must be emphasized that no mechanism has been identified for the PORV to inadvertently open and remain open since its signal to open comes from safety grady eauipment and the Dresser..valves and solenoids are qualified'or an environment in excess of 400'F.

P

4.

Sequence of Events a.

b-C ~

e.

Large steam line break occurs inside containment.

Reactor trip occurs on steam generator low pressure within 5 seconds.

Should the adverse environment cause the PORV to inadvertently open and then remain open, the following steps may also occur.

It should be noted that no mechanism has been identified'hich would cause this to occur.

Steam from PORV fills quench'ank and bursts rupture disk re-leasing steam to the containment and causing additional containment pressurization.

thass removal via PORV causes add'itional void formation within the

'eactor "coolant system;.,

5.

Actions a ~

Short term:

l. Utilities continue t'o investigate qualification levels and location of power cables to PORVs and PORV block valves to assess credibility of this failure mode.

2.

Ensure operators take acti'on to shut PORV and PORV bTock valve if PORV fails open.

Long term:

l.

Complete assessment of PORVs and block valves.

Dependent on the results of that assessment a.

upgrade environmental qualification level of PORVs and block valves; or b.

perform detailed analysis of event if required.

.1

l g /

P

C.

Steam Line Break eolith Feedwater Flow Control S stem Failure 1.

Significant Interaction Effects

. a.

Steam generator filling - causing potential piping structural problems 2.

Assumptions a.

Small steam'-.'linh break inside containment that does not cause an immediate reactor trfp.

I, k

b.

Feedwater flow exceeds steam flow due to failure;of-steam generator level instrument-, indicating flow c.

SAR conservatism.,*

no operator action within'0 minutes 3.

Sequence of Events a.

Small steam line break occurs which does not cause an immediate reactor tnp b.

Steam generator level instrument fails, causing an increase of.

'feedwater flow in excess of steam flow c.

Steam generator begins to fill causing increasedmoisture content.

of steam

d. If no operator action occurs undefi~ed piping structural problems could result e.

It should be emphasized that this event can be prevented by prompt operator action.

Safety grade steam generator level in-strumentation. exists',enabling comparison"with'ontrol grade.

level, instruments.

of-: the. feed,.system; 4.'

Action

, a.

Short term i..

Ensure the operator is aware of this potential interaction'o that he may take prompt corrective action should it occur

t P

b.

Long term Assess the need of upgrading steam generator.

level indication to the feedwater control system ii.

Assess the need to install a safety grade high steam generator level alarm D. 'team Line Break With Failure of Hain Steam Paths Downstream of HSIY's 1.

Significant Interaction Effects a.

Increase post-tri"p return-to-power Wl,

~

2.

Assumptions a.

Large steam'line,break inside containment b.

HSIV on unaffected steam generator fails to close.

- This sequence of events is pertinent only if'his assumption is made.

c.. Downstream of HSIY's main steam paths fail open d.- Initial condition:

full power e.

SAR conservatisms i.

end of cycle core ii.

the most reactive CEA stuck out

iii, steam blowdown through steam line break

,3.

The number of failures which must occur during this event are significant.

First'there must be the large break.

Then the HSIV on the opposite steam generator must fail to close.

There is a stuck rod on reactor trip.

Then steam paths downstream of the HSIY's must be affected.

These include turbine control valves and steam dump and bypass Valves.

The probability of this event occurring is much less than 10 pe~

reactor year.

4.

Sequence of Events.

.a.

. Large steam line break-inside.-containment-'.

Reactor trips on low steam generator pressure trip signal c.

HSIV on unaffected steam generator. fails to close on HSIS d.

Hain steam paths downstream of HSIV open or'fail to close due to control system malfunction caused by adverse environment following large steam line break.

e.

Open main steam paths increase the steam blowdown and increase moderator cooldown effect which adds positive reac-tivity to core.

A post-trip return-to-power is more severe

f g I r

I

under these conditions 5.

Actions a.

Short term should a steam line break occur, ensure operator takes action to isolate all alternate steam flow paths ii.

determine whether this event warrants further consideration, in light of low probability. of all consequential failures which must occur for the. event to. be significant b.

Long term utilities investigate environmental qualification level of the systems involved'i upgrade qualification level of affected equipment if this is determined to be necessary, E.

Steam Line Break lilith Atmos heric Oum Valve Control S stem Failure 1.

Significant Interaction a.

Post-accident controlled cooldown 2.

.Assumptions a.

Steam line break outside containment and upstream of HSIV b.

Atmospheric dump valves on opposite steam line open and remain open~

c.

SAR conservatism no operator action within 30 minutes 3.

Sequence of Events a.

A steam line break. outside of containment but upstream of the'<SIY'ccurs b.

Reactor trip on low-steam-generator-pressure-c.

Atmospheric dump valves upstream of t/SIV's open and remain open due to control system failure

• The failure mechanism identified is a failure of the input signals that would cause the valve to open if operating in the automatic mode.

Although no operator action is assumed for 30 minutes prompt operator action to shut the open valve would mitigate any effects of this event.

/

I

~

d. If no operator action takes place there would be the potential for dry-out and depressurization of both steam generators e.

Failure to shut atmospheric dump valves could inhibit a con-trolled plant cooldown by limiting the ability of the auxiliary feed pumps to deliver to the steam generator(s) 4.

Actions a." Short term opel"ate atmospheric'dump-valves in manual mode, or N

'ii.

ensure operator shuts atmospheric dump valves on=steam line

'ntil control" is assured'.

Long term Continue investi'gation to determine if this failiire mechanism is plausible upgrade atmospheric dump valve control system to viithstand the adverse environment, if'equi red II.

Assessment of Impact of Control System Failures on Feed Line Break Event and CEA Ejection 9

A:

SAR Feed Line Break

'I

'1.

Sequence of Events a.'ain feed line. break occurs do>vnstream of reverse flo>v check valve, discharging main feed and steam generator. fIuid b.

'RCS heatup due to loss of subcooled feed floin

c. 'eactor trip occurs on steam generator lout stater level or high pressurizer-pressure.,-.

Turbine:.trip occurs= on.=

reactor trip I

E d.

- Rapid RCS" heatup=-and'-.pressurization due'- loss of; heat.-".transfer as the ruptured'- steam. generator empties e.

Depressurization of the'ruptured steam generator initiates MSIS and isolates the intact generator f.. RCS pressurization terminates arith opening of primary relief/

safety valves and decreasing core heat flux g.

RCS cooldoivn begins, controlled by the main steam safety valves h.

Auxiliary feed is initiated automatically or by operator action

B.

Feed Line Break With RCS Inventory Control Failure 1.

Significant Interaction Effect a.

Increased RCS pressurization due to liquid filled pressurizer 2.

Assumptions a.

Small feed line break'nside containment b.

Adverse environment'mpacts pressurizer level.instrument causing

"..: indication'to faiT: lowwhich causes thecontrol system to increase.

inventory '(and pressurizer TeveT}'~

c.'nitial conditions i.

102% power

• ii.

steam.bypass control system in manual mode iii. beginning-of-cycle core parameters d.

Analysis conservatisms J

no operator action for at least 30 minutes ii.

no credit for steam generator low water level trip in ruptured unit until empty iii.

heat transfer in ruptured steam generator instantaneously terminated on emptying iv.

failure of the feed line reverse flow check valve, if the break occurs upstream of'he valve 3.

Sequence of Events a.

Feed line break in containment b.

lhain feed spillsfrom break c.

Adverse containment. environment'auses pressurizer level indication to fa'il low caus>.ng RCS inventory to increase d.

Reactor trip occurs on steam generator low water level or high

'ressurizer pressure.

Turbine trips on reactor trip e.

RCS heatup results from rapid decrease in SG heat transfer due to loss of fluid from the ruptured steam generator f.

Pressurizer relief and/or safety valves open

g.

Potential for pressurizer to fill with liquid exists due to high level in pressurizer prior to heatup.

Relief/safety valve relief capacity reduced by liquid discharge 1

h.

Extent of increased RCS pressurization is dependent on time of pressurizer filling relative to the rapid heatup

=4.

Actions a.

Short term alert operator to.this potential failure mode, so that prompt corrective action can be.taken b.

'Long term i.

Perform plant specific analyses to determine upper limit

, allowable for pressurizer TeveT which is consistent with the maximum rate'f level increase and the maximum RCS expansion during the potentially rapid heatup associated with feed line breaks.

ii.

upgrade pressurizer level instrumentation C.

Feed Line Break With PORV Control Failure 1.

Significant Interaction Effects a.

A failed open PORV in combination with a feed line break has not been analyzed 2.

Assumptions a.

Feed line break inside containment b.

PORV's inadvertently open and remain open c.

PORV black valve also fails to close when required d.

No operator action until 20 minutes 3.

PORV would not be expected to remain malfunction since Dresser valves and temperatures in excess of-400oF..

4.

Sequence of Events open due to. actuation solenoids-are qualified for a.

Feed'line break occurs inside containment b.

Steam generator fluid and/or main feed spill from break c.

RCS heatup and pressurizatio'n results from loss of feed flow d.

PORV opens on high pressure and fails to reclose due te adverse environment

~

~I

0 e.

Reactor trip occurs on high pressurizer pressure.

Turbine trips on reactor trip f.

RCS depressurization occurs if PORV's fail to reclose g.

Mass removal via PORV causes void formation within RCS h.

Feed line break in combination with a failed open PORY has not been analyzed 5.

Actions a.

Short term

i. utilities investigate qualification level and location of

'ower cables to PORV's and PORV block. valves to assess

-- credibili'ty of this failure mode ii.

ensure operators take actions to shut PORV 's and PORV block

valves, should this failure occur b.

Long term i.

Complete assessment of PORY's and block valves.

Dependent on results of that assessment A.

upgrade environmental qualification level of PORV's and block valves, or B.

perform detailed analysis of event, if required D.

Feed Line Break 'Hith Feedwater Control Failure 1.

Significant Interaction Effects a.

Overfilling of. the steam generator(s) causing potential structural problems 2.

Assumptions a.

Small feed line break inside containment b.

Feed control in automatic mode C.

Adverse envinnment causes steam. generator level indication to fail low which causes the, feed control system to inGrease feed flow above the steam flow-d.

Ho operator action for 30 minutes

~ /

e

~ ~

3.

Sequence of Events a.

A small feed line break occurs inside containment b.

blain feed spills from break c.

Steam generator level. instrument fails indicating low and causes increased feed flow in excess of steam flow d.

Steam generator begins to fill causing increased moisture content

..--- o.f steam t

e.

If'- no o'perator action occurs undefined structural problems could resul t -.

f. It should be emphasized that this event can be prevented by prompt operator action.

Safety grade level instrumentation exists to compare to control grade instruments.

The feed system can then be controlled manually 4.. Actions a.

Short term i.

ensure the operator

.is aware of the potential failure mode so that he may take prompt corrective action, should it occur b.

Long term assess the need of'pgrading steam generator level indica-tion to the feedwater control system assess the need to install safety grade high steam generator level ala'rm, E.

Feed line Break with Atmospheric Steam Dump 'Control Failure 1.

Significant Interaction Effects a.

Controlled plant cool'down F

2.

Assumptions a.

Feed line break outside containment flow check valve i

and downstream'f reverse b.

Adverse environment impacts.the atmospheric steam dump.control on unaffected steam generator causing an uncontrolled steam release. upstream of the HSIV's c.

No operator action until 30 minutes.*

• The failure mechanism identified is a failure of the input signals that would cause the valve to open if operating in the automatic mode.

Although no operator action is assumed for 30 minutes, prompt operator action to shut the open valve would mitigate any effects of this event.

I 1

3.

Sequence of E s

a.

Feed line break occurs outside of containment downstream of check valve b.

Steam generator fluid and/or main feed spill from break c.

Reactor trip.occurs on steam generator low water level or high pressurizer pressure.

Turbine trip occurs on reactor trip d.

Steam generator pressure increases following turbine trip e.

'Environment could cause atmospheric dump valves upstream of HSIV in unaffected steam generator to open and remain open

f. if no operator action takes place there would be a potential for dry out and depressurization of both steam generators g.

Depressurization of both steam generators may limit the ability of the auxiliary feed pumps to deliver to the steam

'enerator(s).

4.

Actions a.

Short term i.

operate atmospheric steam, dump valves in the manual mode, or ii.

ensure that the operator is aware of this potential inter-action so that prompt corrective action can be taken b.

Long term 1

continue investigation to determine if this failure mechanism ss plausible upgrade atmospheric dump valve control system environmental qualification if required

I 4 J

Potential Effect of Reactor Regulating System During High Energy Pipe Break Events A.

CEA position malfunctions due to steam and feedline breaks and CEA ejection 1.

Significant interaction effect:

a.. Potentially higher reactor power levels prior to reactor trip than presently analyzed

/

2.

Assumptions Small'igh energy pipe break inside containment b.

Reactor regulating system in automatic'mode c.

Adverse environment results in a, low indicated power level from the ex-core sens'or input to the Reactor Regulating.'ystem causing CEAs to be withdrawn F.

3.

Sequence of events a.

High energy pipe break inside.containment of a small enough size where iowediate reactor trip does not occur b.

Control grade ex-core sensor indication fails low due to adverse environmental impact c.

Reactor regulating system causes CEAs to be withdrawn d.

Reactor power exceeds the power previously assumed during the transient e.

Reactor trip occurs due to high energy pipe break at conditions not considered in present analyses 4.

Actions

. a.

Short term

.i.-- place the cont'rol element drive'ystem in manual.

'Y

'- ii. 'iodify emergency. procedures-to'state that the operator should'ot'ake any control action based upon reactor power as measured by the control grade ex-core detectors during high energy pipe breaks b ~

Long term evaluate the consequences of small high energy pipe breaks in containment with CEA withdrawal, if required if required, upgrade the environmental qualification level of the control grade excore detector system

B.

Small Break LOCA With CEA Control System Halfunction 1.

Significant interaction effects Potential exists for increasing power.

This would cause pressure to remain above low pressurizer pressure trip for a longer period than previously assumed 2.

Assumptions a.

Small break LOCA., insi'de containment b.

CEA control system in automatic mode c.

Adverse environment impacts CEA control system or related sensors.

resulting in consequential failure d.

Control system causes CEA to withdraw e.

Standard LOCA licensing assumptions 3.

Sequence of events a.

Small break LOCA occurs inside containment b.

CEA control system in automatic mode c.

Adverse environment caused by rupture potentially causes excore power indication to indicate low power level d.

Should CEAs begin to withdraw, the magnitude of the overpower excursion prior to trip would be increased.

This could produce a higher primary system pressure which cou'1d then delay reactor trip and SEAS and result in higher peak clad temperatures 0.

Action a.

Short term Place the control element drive system in manual'Iodify emergency procedures to state that, the operator should not take. any control action based upon. reactor power.

as measured by the control grade excore detectors during a

LOCA.

b.

Long term Evaluate the consequences of a small break LOCA with CEA withdrawal, and if required upgrade the environmental qualification level

.of the control grade excore instrumentation

~

tP.

t~ w

')

if

1 4i APPENDIX B Re:

St.

Luci e Unit 1

Docket No. 50-335 Safety/Control Interactions POTENTIAL IMPACT ON CONTROL SYSTEflS

'- ".--;..DUE TO ADVERSE Ef'lVIRONMENT RESULTING FROM'HIGH ENERGY LINE BREAKS PRESSURIZER LEVEL CONTPOL SYSTEtl PRESSURIZER PORV CONTROL SYSTEfl REACTOR REGULATING SYSTEM FEEDWATER REGULATING SYSTEM TURBINE CONTROL SYSTEM STEAMi BYPASS SYSTEtl ATMOSPHERIC DUMiP VALVE CONTROL SYSTEM As the CE evaluation was generic, Florida Power 5 Light Company has reviewed the CE document for applicability to the St. Lucie Unit 1 design.

The current status of this review follows.

Evaluation will conti nue as described in Appendix A.

PRESSURIZER LEVEL COiNTROL SYSTEM The CE concern postulates.'the-failure of a pressurizer level instrument in the control..system, which:; in the= absence of operator action, causes;..

the pressurizer to fill', thereby,

a1.lapwing the reactor, coolant systen to go solid.

As discussed in our response--to IE Bulletin-79-01, the-level instruments are -post-LOCA qualified-We therefore do not believe there is a concern in th'.s area.

PRESSURIZER PORV CONTROL SYSTEM CE has not identified a failure mechanism relative to this concern.

Furthermore, this system provides input to the Reactor Protective System

and, as such, is safety-grade and post-LOC/\\ qualified.

We do not believe this item is applicable to St. Lucie Unit 1.

C As a result of IE Information Notice 79-29'nd subsequent letters sent to all

= operating light water reactors under 10,CFR 50.54 (f), Combustion Engineering (CE) has performed a generic evaluation of control systems which, due to the environm'ental effects associated with a high energy line break, could possibly interact with safety systems in an adverse manner.

The results of the evaluation conclude that in no instance was CE able to identify any system interaction which could constitute a substantial safety hazard.

CE did, however, identify seven systems'- warranting additional consideration.

These'ystems are discussed in Appendix A and are listed below:

~p Vy

REACTOR REGULATING SYSTEM RRS The CE concern regarding control rod withdrawal with the Reactor Regulating System (RRS) in automatic control is considered valid.

However, in 1978 a Plant Change/iModification was completed which removed the automatic withdrawal function from the RRS.

Therefore this concern does not apply to St. Lucie Unit l.

FEEDMATER REGULATING SYSTEM F'l1RS The concern in this area assumes a failure in a steam generator level instrument causing the FHRS to supply feedwater in excess of steam

demand, thereby filling the affected steam generator potentially leading to excessive moisture carryover.

The St. Lucie Unit 1 design incorporates a safety grade design feature that automatically closes the feedwater regulating valves at the 88/ steam generator level and trips the reactor at the 90T level.

1!e therefore conclude that this concern is not applicable to St Lucie Unit 1.

TURBINE CONTPOL SYSTEh AND STERl BYPASS SYSTBi As discussed in the CE evaluation (Appendix A) this concern is not considered credible on the basis of the extreme low probability of the sequence of consequential failures required for the scenarios.

ATMOSPHERIC DUtlP VALVE CONTROL SYSTEV The atmospheric dump valves are located upstream of the main steam isolation valves at St. Lucie Unit I, and the postulated failure in this area would be a valid concern were the system to be in the automatic mode during power operations.

However, consistent with the analyses in the FSAR, this system is maintained in the manual mode during normal operations.

l<e believe this method of operation adequately addresses any concern in thi s area.

t.ir 4 i Cy g

STATE Ol"'LORXDA

)

)

COMi TY OP DADE c

)

ss E-A. AQomat,

,'eing first duly sworn, deposes and says:

That he is Execugi~e pic P

~of,F3.orida Power 6

Light Company, the Licensee herein; That he has executed the foregoing document;*that the state-ments maQe in this saiQ document: are true and correct to the best of.his knowledge, information, and belief, and that he is authorized to execute the docum nt. on behalf of aid Licensee E

A Adomat Subscr'ibed and sworn to before me this NOTA'UBLIC, xn an or the County of Dade, Stat of Florida Ily commission expires:

~NQL aav nsu a<am 4~ <0~'~"

~

>os>

4QLnnSStOW 6,<r4~ ~~~,"~

ma~~~

QBU MAYtfAN

1 I

n j'g