ML19329D625
| ML19329D625 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/07/1975 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Rodgers J FLORIDA POWER CORP. |
| Shared Package | |
| ML19329D626 | List: |
| References | |
| NUDOCS 8003160248 | |
| Download: ML19329D625 (4) | |
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Doc, net No.:
50-302 VAMoore FSchroeder LWR TCs ELD IE (3)
Florida Power Corporation LEngle ATU:
Mr. J. T. Rodgers EGoulbourne Assistant Vice President TR BCs
& Huclear Project Mana;;er LWR BCs P. O. Box 14042 ACRS (14) w/ encl.
St. Petersburg, Florida 33733 JMazetis Gentlemen:
B&W Topical Report No. 10103 is presently scheduled to be cubnitted en July 9,1975 for our review in support of your application to construct and operate the Crystal River, Unit 3 facility.
To coiplete the review of your application with regard to compliance with 10 CFR 50 l.6, certain ratorial in addition to that subtitted in the referenced topical report is needed.
Attachnent 1 to this 1cteer is an overall requirceento statemcat delineating all information necessary for the staff to complete its review of ECC.'s capa'oility on each and every application docket.
Ecch USSS vendor (including B&W) has already been provided with all the attcched information except the first two pages.
We urge you to evaluate these requirements and be assured that your submittals on the Crystal River, Unit 3 docket include all the required information outlined in Attachment 1.
Please advise this office uithin 10 days of your schedule for subnitting additional information cs required.
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U A._Schwencer, Chief Light Water Reactors tranch 2-3 Division of Reactor Licenaing Required Ir. formation ec: See ne::t page
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REQUIRF9 INFOR"ATION 1.
Break Spectru= and Partial Loop Operation _
The infor=ation provided for each plant shall comply with the provisions of :he attached =ccorandu= entitled, " Mini =u= Require =ents for ECCS 3reak Spectru: Sub=it:als."
2.
Potential Baron ?recipitation (?UR's Oniv)
The ECCS syste= in each plant should be evaluated by the applican:
(or licensee) to show that significant changes in che=ical concentrations will not occur during the long ter= after a loss-of-coolant accident (LOCA) and these po:en:ial changes have been specifically addressed by appropriate operating procedures.
Accordinely, the applicant should review the syste= capabilities and opera.ng procedures to assure that boron precipi:a: ion would not compromise icng-ter: core cooling capability following a LCCA.
This review should consider all aspects of the specific plant design, including co=ponent qualifica:icn in :he LOCA environment in addition to a detailed review of opera:ing procedures.
The applicant should exa=ine the vulnerability of the specific plant design to single failures that would result in any significant boron precipi:ation.
3.
Single Failure Analysis A singic failure evalua: ion of :he ECCS should be provided by the applicant (or licensee) for his specific plant design, cs required by Appendix K to 10 CFR 50 See:ica I.D.1.
In perfor=in;; this evaluation, the effects of a single failure or operator error that causes any manually controlled, electrically-operated valve :c =ove to a positica :ba: could adversely affec: the ECCS nus; be considered.
Therefore, if this consid-cration has not been specifically reported in the past, the applicants I
upcomin; sub=ittal cus: address this considera:icn.
Include a list of all of the ECCS valves that are curren:ly required by the plant Technical Specifications to have power disconnected, and anv procosed plant modifications and changes to the Technical Specificaticas that =1;ht be required in order to protect against any loss of safety function caused by this type of failure.
A copy of Stanch Technical Position CIC53 13 fro = the U.S. Nucletr Regulatory Cc==ission's Standard Review ?lan is attached to provide you with guidance.
1 The single failure evale: tion should include the potential for passive failures of fluid ys:c=s during long tern cooling following a LOCA as well as single failures of active co=ponents.
For FWR plants,
.the single failure analysis is to consider the potencial boren concentra-
)
proble= as an integral part of long ter= cooling.
4.
Sub= creed Valves I
The applicant should review the specific equipment arrangc=ent with-(
in his plant to deter =ine if any valve =otors within contain=ent will J
become submerged following a LOCA.
The review should include all valve i
motors that may beco=e submerged, not only those in the safety injection syste=.
Valves in other systems may oc needed to li=it boric acid con-
)
centration in the reactor vessel during long ter= cooling or =ay be required for contain=cnt isolation.
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The applicant (or. licensee) is to provide he following information, for each plant:
(1)
Whether or not any valve cotors will be sub=erged following a LOCA in the plant being reviewed.
(2)
If any valve motors will be flooded in their plant, the applicant (or
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licensee) is to:
(a)
Identify the valves that will be submerged.
(b)
Evaluate the poten:ial consequences of flooding of the valves for both the shor: tern and long ter: ECCS functions and containment isols:lon.
The long ter should consider the potential proble of excessive concentrations of boric acid in PWR's.
(c)
Propose a interi: solu:icn while necessary modifications are being designed and i:plemen:ed.
(currently operating plants only).
(d)
Propose design changes to solve the potential flooding proble=.
5.
Containment Pressure ( ?'.iR ' s Un iv )
The ccatain ent pressure used to evalEste tha performance capability of the ECCS shall be calculated in accordance with the provisions of Branch Technical Position CS3 c-1, which is enclosed.
6.
Lou ECCS Reflood Eate (Westinghouse NSSS Only)
Plants that have a Westinghouse nuclear stea supply shall perform their ECC3 analyses utilicin; the proper version of the evalua: ion codel, as defined below:
(1)
The December 25, 1974 versien of the Westin; house evaluation
- model, i.e., the versica without the codifica:icns described in WCAP-8471 is acceptable for previously analv ed plants for which the peak clad temperature turnaround was identified prior to the reflood rate decreasing below 1.1 inches per second or for which the reflood rate was identified to remain above 1.0 inch per second; condi:icas for ehich the December 25, 1974 and March 15, 1975 versicas would be equivalent.
(2)
The March 15, 1975 version of the Westinghouse evaluation n'odel is an acceptable nodel to be used for all previously analyced plants for uhich the peak clad :caperature turnaround was identi-fied to occur after the reflood rate decreased belcw 1.1 inches per cecond, and for '<hich steam coolitig conditions (reflood rate less than 1 inch per second) exist. prior to the time of peak clad te=perature turnaround.
The March 15, 1975 version will be used for all future plant analyses.
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ntR 2 5 1975 MINIPUM REQUIREMENTS FOR ECCS BREAK SPECTRUM SUEMIT*ALS I.
INTRODUCTION The following outline shall be used as a guideline in the evaluation of LOCA break spectrc= submittals.
These guidelines have been for=ulated for conte =porary reactor designs only and =ust be re-assessed when new reactor concepts are sub=it:ed.
The curren ECCS Acceptance Criteria recuires that ECCS cooling perfor:ance be calculated in accordance wi:h an accep:able evaluation model and for a nu=ber of postula:ed loss-of-coolant accidents of dif ferent sices, locaticas and other preperties sufficient to trovide assurance : hat the en: ire spectru=.
of postulated less-of-ccolant accidents is ccvered.
In additica, :he calculation is :o be concucced wita at least three values of a discharge (C ) applied to :he postulated break area, these values spanning coefficica:
D the range fros 0.6 to 1.0.
Sections IIA and IIIA define the acceptable break spectru: for cost operating plants which have received Safety Crders.
Sections II3 and IIIE define the break spectru= require ents for mos: CF and OL case work (excepticas noted later).
Sections IIC and IIIC previde an outline of the =ini u= require:en:s for an acceptabla cotale:e break spectrc=.
Such a cc plete break spectru:
could be apprcpr:ately referenced by some plants.
Sections III3 and IIII provide the excep; ions to certain plant types noted above.
A plant due to reload a portien of its core will have previously sub=itted all ;
or part of a break spectru analysis (either by reference or by specific 1
calcula:icns).
If it is the intention of the Licensee to replace expanded I
fuel with new fuel of the sa:e design (no rechanical design differences which could affect thereal and hydraulic performance), and if the Licensee in: ends to operate :he reloaded core in compliance with previously approved Technical Specificaticns, no additional calculations are required.
If the reload core design has changed, the Licensee shall adop: either of Sections IIA or IIC, or of Secticas IIIA_or IIIC cf this documen:, as appropria:e to the plant type (BWR or PWR).
The criterien for establishing whether paragraph A or C shall be satisfied will be determined on the basis of whether the Licensee can de:enstrate that the shape of the PCT versus break sice curve has not been codified as a consequence of changes to the reload core design.
When the reload is supplied by a source other than the NSSS supplier, the break spectru= analyses specified by Sections IIC or IIIC shall be subsitted as a minica: (as appropria:e to the plant type, SWR or PWR).
Additional sensitivity studies may be required to assess the sensitivity of fuel changes in such areas- ]
as single failures and reactor coolant pu=p perfo igcqe c3 cg-q II.
PRESSURIZED WATER REACTORS
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Operating Reactor Reanalvses (Plants for which Safety Orders were issued)
If calculational changes
- were made to the LSM** to make it wholly in j
- Calculational changes /Model changes--those revisions made to calculational techniques or fixed parameters used for the referenced co=plete spectrum.
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- LBM--Large Break Model; SEM--Scall Break Model
conformance with 10CFR50, Appendir. X, the following minimum nu=ber of breah sizes should be reanalyzed.
Each sensitivity study performed during :he development of the UCCS evaluation rodel shall be individually verified as
. remaining applicable, or shall be repeated.
A plant may reference a break spectrum analysis conducted on another plant if it is the same configuratic-and core design.
1.
If the lar2 cst break si:e results in the hiehest PCT:
a.
Reanalyze the li=iting break.
b.
Reanalyze two smaller breaks in the large break region.
2.
If the lar2est break si:e does net result in the hizhest PCT:
a.
Reanalyze the li=iting break.
b.
Reanaly:e a break larger and a break smaller than the liciting break.
If the liniting break is outside the range of Moody multipliers of 0.6 to 1.0 (i.e.,
less than 0.6), then the li= icing break plus two larger breaks must be analyzed.
If calculaticnal changes have been made,to the S3M to ake it wholly in conferrance with 10CFR50, Appendix n, the analysis of the worst smallbrea)
(SBM) as previcusly determined from paragraph C belcw should gj B.
New CP and OL Case "Jork A complete break spectru= should be previded in accordance with paragraph C below, except for the following:
l.
If a new plant is of the same general design as the plant used as a basis for a referenced cc plete spectrum analysis, but operating parameters have changed wnich would increase PCT or cetal-water reaction, or approved calculational changes resul:ing in mere than 20 F change in PCT have been cade to the ECCS :odel used for the referenced complete spectru, the analyses of paragraph A above should be provide pins a mini =ue of three small breaks (53M), ene of which is the transition.breas."
.ne shape of the break spectru: in the referenced analysis should be justified as remaining applicable, including ;he sensitivity studies used as a bcsis for the ECCS evaluation =odel.
2.
If a new plant (configuration and core design) is appl'icable to all generic s:udies beccuse i: is the sa:e with respect te the generic plant design and parameters used as a basis for a referenced cc:ple:e spectrum defined in paragraph C, and no calculational changes resulting in = ore than 20 F change in PCT were made to the ECCS : del used for
.the referenced complete spectrum, then no new spectrum analyses are required.
The new plant =ay instead reference the applicable analysis.'
- Transition 3rcak (TB)--that break size which is analy:ed with both the LBM and SEM.
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C.
Minimum Recuirenents for a comolete Break Soectrun Since it is expected that applicants will prefer to reference an applicable complete break spectrum previously conducted on another plant, this paragraph defines the -i-i-- np ber of breaks required for an acceptable complete break spectrun analysis, assuming the cold leg pump discharge is established as the vorst break location.
The worst single failure and worst-case reactor coolant pump status (running or tripped) shall be established u:ilizing apprcpriate sensi:1vity studies.
These studies should show tha: the worst single failure has been jus:ified as a function of break size.
Each sensitivity study published during the developmen:
of the ECCS evaluation model shall be individually jus:ified as re=aining applicable, or shall be repeated. Also, a proposal for par:ial icop operation shall be supported by identifying and analyzing rhe worst break size and loca:ica (i.e., idle loop versus operating loop).
In addition, sufficient justification shall be provided to conclude : hat the shape of the PCT versus 3reak Si:e curve would not be significantly al:ered by the partial loop confi;ura: ion.
Unless this information is provided, plant Technical Specifications shall not permit operation with one or core idle reactor coolant pumps.
It cust be denonstra:cd that the containment design used for the break spectrum analysis is appropriate for the specific plant analyzed.
It 1
should be nc:ed tha: this analysis is :o be perfor:ed with an appr,ved evaluation todel wholly in conformance with the curren: ECCS Acceptance Criteria.
1.
LBM--Cold Leg-Eeactor Coolant Pu=p Discharge a.
Three guillotine type breaks spanning at least the range of Moody nultipliers between 0.6 and 1.0.
b.
One split type break equivalen: in size to twice :he pipe cross-sectional area.
e c.
Two intermediate split type-brsa'is, s d.
The large-break /small-break transition split.
2.
LBM-* Cold Leg-Reactor Coolant Pu=p Suction Analyze the largest break size fro part 1 above.
If the analyses in part 1 above shculd indica:e that the worst cold leg break is an intermediate break size, then the largest break in the pu=p suction should be analyzed with an explanation of why the sa:e trend would i
not apply.
3.
LBM--Hot Leg Piping Analyze the largest rupture in the hot leg piping.
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4.
SB'i-- Split s Analyze five different s=all break sizes.
One of these breaks must include the transition split break.
The CFT line break nus be analyzed for B&W plants.
This break =ay also be one of the five s=all breaks.
III.
BOILING WATER REACTCRS The generic nodel developed by General' Electric for SWRs propesed that split and guillotine type breaks are equivalen: in determining bicudewn phen:cena.
The staff cencluded this was acceptable and : hat the break area =ay be considered at the vessel no::lc with a zero 1 css coefficient using a two phase critical flev odel.
Changes in the break area are equivalent to changes in the Moody =ultiplier.
The mini un number of breaks required for a cerclete break spectru analysis, assuning a su::ica side recirculation line break is :he design basis accident (DBA) and the vers single failure has been established utili:ing appropriate sensitivity studies, are shewn in paragraph C below.
Also, a propcsal for partial icap operatien shall be supported by iden:ifyi?.g and analyzing the vers:
break 5.:e and loca:ica (i.e., idle loop versus operating leop).
In additica, sufficient justifica:ica shall be provided to conclude that the shape of the PCT versus Ereak 31:e curve would not ba si;nificantly altered by the partial loop configura:icn.
Unless this infor:a:icn is provided, plant Technical Specifica:icns shall not permit operatica with one er ore idle reactor coolant pumps.
A.
BWR2, SWR 3. and 2WEa Reanalysis (Plancs fer which Safety Orders were issued)
If the referenced lead plant analysis is in accordance with Section III, paragraph C belev, the fellowing.inimun number of break si:es should be reanalyzed.
It is to be no:ed tha: tha lead plan: analysis is to be perforced with an approved evaluation model wholly in cenf or:ance with the curren: ECCS Acceptance Criteria.
A plant cay eference a break r
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=2ti
spectru: analysis conducted en another plar Y5
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j"b and core design.
g Each sensitivity s:udy published during the develop =ent of the ECCS evaluation : del shall be individually justified as remaining applicable, or shall be repeated.
1.
If the larces: break results in the hiehes: PCT:
a.
Reanaly:e the limiting break with the app cpriate referenced single failure.
b.
Reanaly:e the worst s=all break with the' appropriate referenced single failure.
c.
Reanaly:e the transicion break with the single failure and =cdel that predicts the. highest PCT.
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i 2.
If the largest break does not result in the hiehest PCT:
Reanalyze the limiting break, the largest break, and a smaller break.
a.
If calculational changes have been made to the SBM to :-ke it wholly in confor=ance with 10CFR50, Appendix K, reanaly:e the small break (53M) in accordance with Section IIIC.
B.
New CP and OL Case Work A co=plete break spectru: should be provided in accordance with Section III, paragraph C below, except for the following:
1.
If a new plant is of the same general design as the plant used as a basis for the lead plant analysis, but operating parace:ers have changed which would increase PCT or =etal-wa:er rea:: ion, or approved calculational changes have been =ade to the ECCS :odel resulting in
= ore :han 20 F cnange in FCT, :he analyses of Sectica III, paragraph A above should be provided plus a =inicu= of three stall breaks (S5M),
one of which is the transi: ion break.
The shape of the break spectru=
in the lead plant analysis should be justified as re=sining applicable, including the sensitivity s:udies used as a basis for the ECCS evaluation =odel.
2.
If a new plan: (configura: ion or core design) is applicable to all generic studies because i: is :he same vi:h respect to the generic plant design and parameters used as a basis for a referenced co=plete spectru= defined in paragraph C, and no calculational changes resulting in more than 200? change in PCT were made to the ICCS codel used for the referenced co=ple:e spectru=, then no new spectru= analyses are required.
The new plant may ins:ead reference thc applicable analysis.
C.
Mini =un Recuire=ents for a Complete Ercak See: run This paragraph defines the ninimum nu=ber of breaks required for an acceptable complete spectrum analysis.
This complete spectrum analysis is required for each of the lead plants of a given class (SWR 2, SWR 3, BWR;,
BWRS, and SWR 6).
Each sensitivity study published during the develop =ent of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated.
1.
Four recirculation line breaks at the wors: location (pu=p section or discharge), using the LSM, covering the range f rom the transitica break (T3) to the D3A, including C3 coefficients of fro: 0.6 to 1.0 times the D3A.
2.
Five recirculation line breaks, using the SSM, covering the range from the s=allest line break to the T3.
3.
The following break locations assu=ing the worst single d]s a.
largest steamline break I
b.
largest feedwater line break h
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largest core spray line break c.
d.
largest recirculation pump discharge or suction break (opposite l
side of worst location)
D.
BWR4 with "Medified" ECCS Same as Section IIIC.
g, BWR5 Same as Section IIIC.
F.
BWR6 Same as Section IIIC.
IV.
LOCA PA?.05TERS OF INTEREST A.
On each plant and for each break analyzed, the following para =eters (versus' time unless otherwise noted) should be provided on engineering graph paper of a quality to facilitate calculations.
--Peak clad tenperature (ruptured and unruptured node)
--Reactor vessel pressure
--Vessel and downco=er water level (PWR only)
--Water level inside he shroud (SWR only)
--Thermal power
--Containment pressure (PWR only) 3.
For the worst break analyzed, the following additional para eters (versus time unless otherwise noted) should be provided on engineering
~
graph paper of a quality to facilitate calculations.
The worst single failura and worst-case-reactor coolant pu=p status vill have been established utilizing appropriate sensitivity studies.
--Flooding rate (PWR only)
-Core flev (inlet and outlet) 0
--Core inlet enthalpy (37R only)
--Heat transfer coefficients
-MAPLHCR versus Exposure (BWR only)
--Reactor coolant temperature (PWR only)
--Mass. released to containment (PWR only)
--Energy released to containment (PWR only)
,k
-PCT versus Exposure (BkT only)
- Containment condensing heat transfer coefficient (Fk2 only)
--Hot spot flow (Pk2 only)
-Quality (hottest assembly) (Pk7 only)
-Hot pin in:ernal pressure
-Hot spot pellet average temperature
-Fluid temperature (ht: test assembly) (Pk'R only)
C.' A tabulation of peak clad temperature and =etal-water reaction (local and core-wide) shall be provided across the break spectru=.
D.
Safety Analysis Reports (S.Gs) filed with the NRC shall identify on each plot the run ca:e, version number, and version date of the co:puter
- codel utili:cd for :h2 LCCA analysis.
Should differences exist in version nu.ber or version date from the cost current code listings made availabic to the N?.C staf f, then each modification shall be identified with an assessmen: cf 1: pact upon PCT and retal-water reaction (lecal and cere-wide).
E.
A tabulation of ti=es at which significant events occur shall be provided en each plant and for cach break analyzed.
The following events shall be included as a minimun:
-End-of-bypass (Pk2 only)
--Beginning of core recovery (Pk'R only)
-Time of rupture
-Jet pumps uncovered (Sk2 only)
-MCPR (Sk3 only)
--Time of rated spray (Bk2 only)
-Can quench (Sk% only)
-End-of-b wdevn
-Plane of interest uncovery (3k2 only)
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