ML18044A406

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Requests Tech Specs for Auxiliary Feedwater Sys Mods & Revision of Normal & Emergency Operating Procedures.Sample Tech Spec Pages Encl
ML18044A406
Person / Time
Site: Palisades Entergy icon.png
Issue date: 12/21/1979
From: Reid R
Office of Nuclear Reactor Regulation
To: Hoffman D
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
References
TASK-10, TASK-RR NUDOCS 8001110008
Download: ML18044A406 (15)
v • d • e


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Docket No. 50-255-Mr *. 'David P. Hoffman Nu Cl ear L kens frig Ad'mi ni strator * *. * *:.

  • Consumers. P0\\1er Company 212.West Michigan Avenue -.*.

Jackson, Michi*gan. 49201

Dear Mr *. Hoffman:

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SUBJECT:

. -~~[~~~6~~ ~~l~~ATlON *OF *.AU~J_L.IARY_,-~EEDW~TER. SY.ST.EMs;:Ar.

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In recent C01rt111U.ni.cati.on$, your staff has.ind*i~ated that.: a.proposed design, using control grade components. which.would autQmatical1y. initiate the auxiliary feedwater"systellis at your facility upon the ross of main feedwater.f'low will"

. be submitted in the n~ar future*.-'This s*ubmittal*was in resp.onse to Short.: -

. Tenn.Recomme_ndation 2.* 1.7.a, ~/\\uto Iriit-iation of.the.Aux*it1ary Feedwater.,

System", as *c1arifi-ed in our letter.October 30;.1979 which was addressed to all operating *m,iclear**power plants_~_

  • '.l'

~We will review your prcfposed design,~gains:i each.. of t~e s~ven position*s.

stipulated.fn Shorf-Terin.Reconmendation 2.1.7._a.< In response.t;o tnis recorrmendation, some licensees have :raise~ the_iss1;1e of-the appliCability of

'current analysis of"a main."stealT_l.*lin~ pr,epk* *or:_ main' feedwater.11ne break assuming.

  • early. initiation of auxiliary-"feedwater -flow w'ith *a 'failure to limit flow

,tot.he affected steam g~f.!erator.* *1n*quest1oriis whether.toe.change in

. assumptions* would increase the cal culate_d-containment.pressure or the.

likelihood of* return to.power.-* T~ese questions are b"elieved applicable for either manual. or automatic in1tiat.ion-*of the auxiliary.feedwater

  • . system. *You are requested t*o -resol.ve.thi's. concern by submit.ting an

. c.:analysihs *within twenty"(20}" days :after receipt:'of this_,let:ter.

. - (telecop1ed *on date.sign~d). The enclosure. to thts letter* provide~

  • **'a Hst. of.questions_and*.irifonnation you.s.hoqld addressas~approJ)riate.

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. As* a res~1t -~f.this *conce*rh--ri*nd pur$uant*t~o o~~r lette~ ~f October.30,

  • :1979,. you should not. implement.autpmatic;ally ihltjated AFWS flow until
  • '* we have* completed our rey'iew arid issued an approv~l~. H_owever, 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

.adve*~~*y af:fect1 '19* the manual-start AFWS.

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NRC FORM 31.8 (9[76) NRCM 0240
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pocket l'!o. -~o-255

  • Mr~* David P. Hoffman
  • t~uclear Licem~1ng.Administrator*

Consumers Power Coinpany

Jackson, Mich19.an 49201 *

Dear Mr.* Hoffman:

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~ECEMBER ~ 1 1919

SUBJECT:

AUTOMAfIC INITIATION OF AUXILIARY.FEEDWATER-SYSTEMS AT_

.. PALISADES PLAN}

~

In recent communications~* your st~ff.has fndie.a1;ed.that~ proposed.design,:

using control grade*compol'lents,.which would*automatically :initiate the auxiliary feedwater systems at your. facility upon the Toss *of m_afo *fee¢.~ater flow will.,

be submitted *1 n the near.future. *Thf s submfttal was' i_n response. t-0 Short-Term Recommeridat ion 2.1.7.a~ -"Au.to I niti at ion *of* the* Aux11 fary Feedwater *.

System 11, as ~larified tn our letter*Octot>er 30*, 19~9 which was addressed to all operating nuclear_pdvrnr plants~~.:

~~e will review yo~r proposed,,d~s{gn against each* of'the.~seven *positions stipulated in Shoft... Tenn.~Recommendatfon 2.1.1.a *. rn response to this recommendation, s0m~ licensees.have' rqi~ed.tne issu~ of the applkabi:lity of.

current analysis of*a rnain.'steam' 1ine.bre~k or*.ma*in feedwater line break.assuming.*

early initiation of. ~*uxiliary* feedt1ater. flow**with a. -failure to -limit flow*

to the a:ffected stea'm. generator *. il n quest.fon ts whether. the change ln assumptions would increase the.calculated containment pressure or the

.likelihood.*of r~turn to pCMer.

These questions* are believed applicable for. either manual or automatic initia_tion of the *auxiliary feedwater

  • system. **You are requested t.o *resolve **this *concern.by *submitting* an.
  • analyses or evaluations*.within twenty- {20) days after receipt of this*
  • lette*r (telecopied on date slgnect).

The enclos.ure to this letter**

Provided a. list.of. gues~~ions and* infonnation yo1(should _address as*

appropriat,e!'

. *' ~ * :,.:. : '*

  • ... /

You.a~e *requested* to**pr6pose *Technical Spe~1fiCations--f.*o~-lhe 'AfWS;-~ocii ficat1oris.~

Sample Technical Specifications are e"!closect Jar your consideratio*n.

  • In addition,

. ypu wi 11 need to re vi ~e normal and emergency operating '*procedures' as requ1 reci' by this modificati.on and tr~i,l'.l the*plant.operations people,.as,'required by these

  • procedures *. Part.icular atteritiori to the inearis of contrpllfog the.bypass capabi.l1ty of -the automatic.. AFWS turbine*start. si*gnal jsrecormiended *. *

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NRC FORM 318 (9*76).NRCM 0240

--tr~.s. GOVER.~MENT.-PfllNT1N'G,QFFi'cE:\\91~-289*3~9-_.*.....

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.*You are 'request~d to _propose Technical Spec_if1cations for ~he.AFWS modifications_~

Sample Technical Specifications are enclosed for your consideration *. Iri addition, yo1.rwi.1l ne.ed to revise normal and e'inergency* operating procedures as required by this modification and train' the plant operations. people as re.quired by these.

procedures. Part1cular attention to the means of controlling the* bypass.

capa_bility of the automatic ftFWS turbf ne start s1 gnal 1 s_ recommended.

Enclosure:

Sample TS Pages cc: w/encl osure.

See next page..

Distribution:

Docket files

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--r Dennis r~* Zifimann;,Chief

  • _. _Operating Reactors Branch #2 Qiv1~ion 9f qperatfng ~eac~or~

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UNITED ST ATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Docket No. 50*255 Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201

Dear Mr. Hoffman:

DECEMBER 2 l G

SUBJECT:

AUTO~ATIC INITIATION OF AUXILIARY FEEDWATER SYSTEMS AT PALISADES PLANT In recent col'!T!lunications, 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-Tern Recorrmendation 2~1.7.a, "Auto Initiation of the Auxiliary Feedwater Systen", as clarified in our letter October 30, 1979 which was addressed to all operating nuclear pCMer plants

  • We will review your proposed design against each of the seven ~,sitions stipulated in Short-Term Reconmendation 2.1.7.a.

In response to this reconmendation, 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 fl<M to the affected steam generator.

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 concern and pursuant to our letter of October 30, 1979, you should not implement automatically initiated AFWS flCM until we have completed our review and issued an approval.

HCMever, 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 AFWS *.

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

Sar.ple Technical Specifications are enclosed for your consideration.

In addition, ya.i will need to revise nonnal 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 autanatic AFWS turbine start signal is recorrmended.

En::l osure:

Sample TS Pages cc:

w/encl osure See next page Sincerely, j;r'}

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Dennis L.

ZiemaOR~ Chief Operating Reactors Branch #2 Di vision of Operating Reactors

Mr. David P *. Hoffman cc M. I. Miller, Esquire Isham, Lincoln & Beale Suite 4200 One First National Plaza Chicago, Illinois 60670 Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Myron M. Cherry, Esq~ire Suite 4501 On~ IBM Plaza Chicago, Illinois 60611 Ms. Mary P. Sinclair Great Lakes Energy Alliance 5711 Summerset Drive Midl~nd, Michigan 48640 Kalamazoo Public Library 315 South Rose Street Kalamazoo, Michigan 49006 Township Supervisor Covert Township Route 1, Box 10 Van Buren County, Michigan 49043 Office of the Governor (2)

Room 1 - Capitol Building Lansing, Michigan 48913 Di rector, Technicq.l Assessment

.Division Office of Radiation Programs *

(AW-459) u~ s. Environraental Prdtection

._gency Crysta 1 ~:a 11 #2 Arlington, Virginia 204~0 U. s. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:

EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604 Charles Bechhoefer~ Esq~, Chairman Atomic Safety and Licensing Board Panel U. s. Nucl:::ar Regulatory Commission Washingtor, D. C.

20555 Dr. George C. Anderson Department of Oceanography University of Washington Seattle, Washington 98195 Dr. M. Stanley Livingston 1005 Calle Largo Santa Fe, New Mexico 87501

Enclosure e

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

Return to Power

1. Provide the results.-0f analyses of main ste~m line breaks that are the most limiting with respect to Juel failure resulting from return to power.

Analyses should be presented covering:

a.

Break inside containment b.-

Break outside contai~ment

c. Avaiiahility or loss of offsite power Justify omitting an analysis for any of the above.
2. Provide the time seauence of all actions and events occurrin~

duriniz: each of the postulated stea~ line break transients.

1hese events and actions should include:

~

a.

Reactor scram

b.

Turbine trip

c.

Steam line isolation

.d.

Fee~water isolation

e.

ECCS a.ctuation

f.

. Auxiliary feedwater actuation and control

g.

Safety/relief valve actuation (prilrary an~ secondary

  • systems)
h.

Operator actions (define credit for operator action)

i.

Initiation of onsite power. (f( required).

3.,

For each of the above,* idf>ntify the initiatin[? signal, the**

protection system that, initiates the action, and the extent of the action ending with the time the element (i.e., HSIV, turbine stop, turbine control, turbine bypass,. etc.) reaches its new condttion *. 1he above events are to reflect the expected response. of the plant and systems.

i *1.

4.
5.
6. Identify and justify any equipment that does not meet Regulatory Guides and IEEE-279 requirements.

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

  • presented consider the worst single failures from a fuel

. failure s~ndpoint. Note that. norr-al control systems should not be considered to function if their action would be beneficial With reSpP.Ct to fuel failurP.S.

Provide the followirig information as a function of time:

a.

Minim in DNBR

b.

Cladding temperature if DNBR limit is exceeded

c.

Feed~~ter flow into faulted and nonfaulted stP.am 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* tefllpera ture

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 liave turned around and.the subcri ticali ty marp.in is increasing.

Note the DNBR calcul~tions* must reflect the ini tia.l plant perturbations due to moderator 'and -pres'surP..decrease and loss of off site power (if appropriate).

  • Also discuss how the effects of a* stuck rod are considered when cr:ilculatiniz DNBRs af~er the* rods have been inserterl.

If fuel dar.iage occurs (i.e., violation of DtJBR), provide fraction of fuel that failed and offsite dose calculations.. Also prov1de and justify DNB *correrlations used. in* ttte analyses.

.... B.

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

l~ 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.

b.

c.

d.

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

Provide the basis for this assumed flow rate.

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

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.

Discuss the design provisions in the AFS used to tenninate the AFS flow to the affected steam generator. If operator action is required to perform this function;; discuss the infonnation 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 co~ponents relied upon to limit the mass and.energy release and the containment pressure response.. The single failure analysis should include, but not

. necessarily:be limited to: partial loss of containment cooling.

systems and* failure of the ~FS isolation valve to close.

f. For the single active failure case which results in the maximum containJllent atmosphere pressure, provided a chronology of events.
  • Graphically, show theccintainment atmosphere pressure as a function
  • of time for:at least 30 minutes following.the accident.. *For this

I.

  • to be at the pump run out fl ow (if a run out contra 1 sys tern is
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 *

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FUNC°TIONAL UNIT

9.

EMERGENCY FEEDWATER

a. Manual
b. Steam Generator Level-Low
c. T t?1:dwil tcr Flow~Low d.. ' Steam Generator Pressure-Low **
e. Safety Injection TABLE*3~3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TOTAL NO.

OF-CHANNELS 2 sets of 2

  • per FDW 11ne 4/SG 4/FOW line 4/SG CHANNELS TO TRIP l set of 2 per FDW line 2/SG 2/FOW l 1ne 2/SG MINIMUM c*:ANNELS OPERABLE 2 sets of 2 per FDW 11ne 3/SG 3/FOW l 1ne 3/SG APPLICABLE MODES l
  • 2. 3, 4
1. 2. 3, 4 l
  • 2, 3. 4 l ' 2, 3, 4 (See Safety Injection in1tiat ing functions and requ1 reme*nts)
  • The prov1sions of Spec1f1cat1on 3~0.4 are not app11cable.

ACTION A *9 O*

O*

B*

3.0.4 Entry. into an OPERATIONAL MODE or other specified applicability cor.~ition *

. shall not be made unless the *conditions of the Limiting Conditi-0n for Oper?tion are met without reliance on provisions contained in the ACTION statements :.rnless other-wise excepted. This provision shall not prevent passage through OP~TIONAL MODES as required to comply with.ACTIQN statements.

ACTION STATEMENTS ACTION A -

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERAELE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be 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 nours.

ACTION B -

With the number of OPERABLE channels one less than the Total Number of Channels, opera~ion may proceed provided the follC1io1ing

  • conditions are satisfied:
a.

The inoperable channel is placed in the tripped conditicn within 1 ho.ur.

b.

A 11 functi ona 1 uni ts 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.l.

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TAOLE J.3-4 (Cont1nur.d}

.ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UN IT TRIP.VALUE ALLOWABLE VALUES

9.

EMERGENCY FEEDWATER

" il. Manual Not Applicable Not Appl 1cable

-h. Steam-Genera tor.

. ti Level-Low

c. fccdwatcr Flow gpm gpm

-I.ow

d. Steam Generator psia ps1a Pressure-Low
  • e *. Safety Injection (see Safety Injection Setpo1nts) i,

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TABLE 3.3-5 (Continued) 8\\GINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTIOW RESPONSE TIME IN SECONDS

1. Manual Emergency Feedwater System
  • Not Applicable
2.

Steam Generator Pressure-Low Emergency Feedwater System

3.

Steam Generator Level-Low Emergency Feedwater System

4. *Feedwater Flow-Low Emergency Feedwater System

... */

NOTE:

Response tirr..; for Motor-dri v:n Emergency Feedwater Pumps on a 11 Safety Injection signal starts

  • Diesel generator starting and se~uence loading delays included.
    • Jiesel gener~tor starting and secuence loading delays not included *. Offsite
>ewer avail a:il e.

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ENGINEERED SAFETY FEATURE ACTUATION SYSTEM.INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL CllANNEL CHANNEL FUNCTIONAL MODES IN li:U CH FUNCTIONAL UNIT CHECK CALIBRATION TEST SURVEILLANCE REQUIRED

.. l. EMEHGENCY FEEOWATER

.. a. Manual In1t1at1on N.A *.

  • N.A.

M(*)

.l,2,3,4 e

b. Steam Generator s

R M

l. 2, 3, 4 Level-Low
c. Feedwater
  • s R

M 1

  • 2, 3, 4 Flow-Low
d. Steam Generator s

R M

. 1 I 2, 3, 4 Pressure-Low

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 e associated with manual safeguards actuation shall receive a CHANNEL FUNCTIONAL TEST at least. once per 31 days.

I

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