ML19329C455
| ML19329C455 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/07/1975 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Roe L TOLEDO EDISON CO. |
| References | |
| NUDOCS 8002130879 | |
| Download: ML19329C455 (30) | |
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Y VAMoore Docket No.: 50-346 FSchroeder MMLWR TCs ELD IE (3)
LEngle Toledo Edison Company EGoulbourne ATTN:
Mr. Lowell E. Roe TR BCs Vice President LWR BCs Edison Plaza ACRS (14) w/ enclosure i
300 Madison Avenue JMazetis Toledo, Chio 43652 Gentlemen:
35W Topical Report No.10105 is presently under review in support of your application to construct and operate the Davis-Besse. Unit i facility. To completo the review of your application with regard to compliance with 10 CFR 50.46, certain material in addition to that submitted in the referenced topical report is needed. to this latter is an overall requirements statement delineating all infornation necessary for the staff to complete its review of ECCS capability on each and every application docket. Each NSSS vendor (includin;;
36W) has already been provided with all the attached infor ation except the first two pages.
We urge you to evaluate these requirements and be assured that your submittala on the Davis-3 esse, Unit 1. docket include all the requirad information outlined in Attactusent 1.
Picase advise this offica within 10 days of your schedule for submitting additional information as required.
Sincerely, krN : Ind h.
i A. Schwencer, Chief Light Water Reactors Branch 2-3 Division of Reactor Licensing Required Information ces: See next page x7886/ LWR 2-3 C,-I
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2-Toledo Edison Company w/o enclosure ecs: Donald H. Hauser, Esquire The Cleveland Electric Illuminating Co.
P. O. Box 5000, Room 610 Cleveland, Ohio 44101 Cerald Charnoff, Esquire Shaw, Pitt:::an, Potts, Trowbridge and Madden 910 - 17th Street, N. W.
Washington, D. C.
20006 Leslie Henry, Esquire Fuller, Seney, Henry & Hodge 300 Madisen Avenue Toledo, Ohio 43604 becs:
J. R. Buchanan, ORNL w/o enc 1.
T. B. Abernathy, DTIE w/o enc 1.
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e Attachment. I' REQUIRED INFORMATION 1.
Break Spectrum and Partial Loop Operation _
The information provided for each plant shall comply with the provisions of the at: ached =ecorandum entitled, " Mini =um Requirements for ECCS Break Spectrus Submittals."
2.
Potential Boron Precioitatien (?Un's Oniv)
The ECCS syste: in each plan: should be evaluated by the applicant (or licensee) to show that significant changes in chemical concentrations will not occur during tha icng term after a loss-of-coolan: accident (LOCA) and these potenfial changes have been specifically addressed by appropriate operating procedures. Accordingly, :he applican: should review the system capabilities and operating 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.inclufing ec=ponent qualification in the LOCA environment in addition to a de: ailed review of operating procedures.
The applicant should examine the vulnerability of :he specific plan: design to single failures tha: vould result in any significant boron precipi:ation.
3.
Single Failure Analvsis A sincie failure evaluation of the ECCS should be provided by the applicant (or licensee) for nis specific plan desi;n, cs receired by Appendix K to 10 CFR 50, See:1cn 1.3.1.
In perfor=ing this evaluatica, the effects of a single failure or operator error that causcs any manually controlled, electri: ally-opera:ed valve to teve :o a positica :hc: could adversely affect the ECCS cus; be considered.
Therefore, if this censid-
~:he applicants cration has not been spe:ifically reported in the past, upcoming submi::a1 mus: address this censidera:icn.
Include a list of all of the ECCS valves : hat are curren:ly required by :he plan: Technical Specifications to have power disconnected, and any prooosed plant modifications and changes :o :he Technical Specifica: ions : hat might be required in order to protect against any loss of safety function caused by this type of failure. A copy of Branch Technical Position ::CIS IS from the U.S. Nuclear Regulatory Co==ission's Standard Review Plan is attached to provide you with guidance.
The single failure evslu.-tion should include the potential for passive failures of fluid systems during long torn cooling following a LOCA as well as single failures of active components.
For ?WR plants, the single failure analysis is to consider the potential boron concentra-problem as an in:egral part of long term cooling.
4.
Sub=cr;ed Valves The applicant should review the specific equipment arrangement with-in his plant to determine if any valve =otors wi:hin contain= cat will become submerged following a LOCA. The review should include all valve motors that may become sub=crged, not only those in the safety injection system. Valves in other syste=s may be needed to limit boric acid con-centration in the reactor vessel during long term cooling or =ay be required for containment isolat ion.
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t The applicant (or licensee) is to provide the following infor=ation, fo-each plant:
(1) Whether or no: any valve motors will be sub=erged following a LOCA in the plant being reviewed.
(2)
If any valve =otors will be flooded in their plant, the applican: (or licensee) is to:
(a)
Identify the valves that vill be submerged.
(b)
Evaluate the potential conseque c., of ficoding of the valves for both the short ter= and long cer= ECCS functions and containment isolation.
The long ter= sheuld consider the potential probie= of excessive concentrations of boric acid in PWR's.
(c)
Propose a interi= solution while necessary =odifica: ions are being designed and i=plemented.
(curren:ly operating plants only).
(d) Propose design changes to solve the potential flooding problem.
5.
Containment Pressure (?N2's oniv)
The contain=en7 ressure used to evaluate the performance capability of the ECCS shall be calcula:ed in accordance with the provisions of Branch Technical Fosi:ica C53 6-1, which is enciesed.
6.
Lou ECCS Reflood Race (Westinzhouse NSSS Only)
Plants that have a b*astinghouce nuclear steam supply shall perfor=
their ECC3 analyses utiliaing the proper version of the evaluation model, a'
fefined below:
(1) The Dece=ber 25, 1974 version of the Wes:inchouse evaluation
=odel, i.e., the version without the codifications described in WCAP-8471 is accep:able for previously analyced plants for trhich the peak clad te=perature turnarcund was iden:ified prior to :he reflood rate decreasing below 1.1 inches per second or for which the reflood ra:e was iden:ified to re=ain above 1.0 inch per second; conditicas for which the Dece=ber 25, 1974 and March 15, 1975 versions would be equivalent.
(2) The March 15, 1975 version of the Westinghouse evaluation ='odel is an acceptable model to be used for all previously analy:ed plants for which the peak clad te=perature turnaround was identi-fled to occur after the reflood rate decreased below 1.1. inches per cecond, and for which sten: cooling conditions (reflood rate less than 1 inch per second) exist prior to the time of peak clad temperature turnaround. The March 15, 1975 version will be used for all future plant analyses.
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,s D 2 51975 M75IMUM REQUIREMENTS FOR ECCS BREAX SPECTRtM SUEMIT*ALS I.
INTRODUCTION The following outline shall be used as a guideline in the evaluation of LOCA break spectru= sub=ittals.
These gaidelines have been fot=ulated for conte =porary reactor designs only and =ust be re-assessed when new reactor concepts are sub=icted.
The current ECCS Acceptance Criteria requires ":a: ECCS cooling perfor:ance be alculated in accordance with an acceptable ecaluation =odel and for a nu=ber of postulated loss-of-coolant accidents of dif ferent sizes, locations and other propertica sufficient to orovide assurance that the entire seectre:
of postulated less-of-ccolant accidents is ecvered.
In addition, the calculation is to be concuctec with at least :nree values of a discharge (C ) applied to the postulated break area, these values spanning coefficient D
the range fro = 0.6 to 1.0.
. Sections IIA and IIIA define the acceptable break spectru= for =ost operating plants which have received Safety Orders.
Sections II3 and III3 define the break spectru requirements for nos: CF and OL case work (exceptiens noted later).
Sections IIC and IIIC provide an cu:line of the =ini=u= require =ents for an acceptable corolete break spec::c=.
Such a co=plete break spectru:
could be appropriately rc:erenced by so=e plants.
Sections IIID and IIIE provide the exce):lons to cer: sin plant types notad above..
A plant due to reload a portida of its core will have previously sub=itted all or part of a break spectru= analysis (either by referen'ce or by specific calculations).
If it is the inten: ion of the Licensee to replace expended fuel with new fuel of the sa=e design (no =echanical design diff erences which could affect ther=al and hydraulic perfor=ance), and if the Licensee intends to operate the reloaded core in co=pliance with previously approved Technical Specificaticas, no addi:ional calculations are required.
If the reload core design has changed, the Licensee shall adop: ei:her of Sections IIA er IIC, or of Sections IIIA or IIIC of this docu=ent, as appropriate to the plant type (BWR or ?WR). The criterion for establishing whether paragraph A or C shall be satisfied will be deter =ined on :he basis of whe:her the Licensee can de=onstrate that the shape of the PCT versus break sice curve has not been =odified as a consequence of changes to :he reload core design. When the reload is supplied by a source other :han the NSSS supplier, the break spectru= analyses specified by Sections IIC or IIIC shall be sub=itted as a mini =u= (as appropriate to the plant type, BWR or PWR).
Additicnal sensitivity studies =ay be required to assess the sensi:ivity of fuel changes in such areas as single failures and reactor coolant pu=p perfor=ance.
II.
PRESSURIZED WATER REACTCRS Operatinz Reactor Resnalvses (Plants for which Safety Orders were issued)
A.
If, calculational changes
- were =ade to the LBM** to =ake it wholly in
- Calculational changes /Model changes--those revisions =ade to calculational techniques or fixed parameters use,d tor the referenced co=plete spectru=.
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- LBM--Large Break Model; SBM--S=all Break Model n
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conformance with 10CFR50. Appendix K, the foAlowing minieu= nu=ber of break sizca should be reanalyzed.
Each sensitivity study perfor=cd during the l
development of the ECCS evaluation model shall be individually verified as remaining applicable, or shall be repeated. A plant may reference a break spectru= analysis conducted on another plant if it is the same configuration and core design.
1.
If the largest break size results in the hiehest PCT:
a.
Reanalyze the limiting break, b.
Reanalyze two s= aller breaks.in the large break region.
2.
If the largest break size does not result in the hi: hest PCT:
a.
Reanalyze the 11=iting break.
b.
Reanalyze a break larger and a break
- aller than the li=iting break.
If the li=iting break is outsice the range of Moody multipliers of 0.6 to 1.0 (i.e., less than 0.6), then the li=iting break plus two larger breaks =ust be analyzed.
If calculaticnal changes have been =ade to the S3M to make it wholly in confor ance with 10CFR50, Appendix K, the analysis of the worst s=all break (SBM) as previously deter =ined fro = paragraph C belcw should be repeated.
B.
New CP and CL Case Work A complete break spectru= should be provided in accordance with paragraph C below, except for the folicwing:
1.
If a new plant is of the same general design as the plant used as a basis for a referenced co=plete spectru= analysis, but operating para =eters have changed which would increase FCT or =etal-water reaction, or approved calculational changes resulting in =cre than 2007 change in PCT have been =ade to the ECCS =cdel used for the referenced co=plete spectru=, the analyses of paragraph A above sho,uld be provided plus a =ini=u= of three s=all breaks (53M), one of which is the transition break.= The shape of the break spectru= in the referenced analysis should be justifiad as re=sining applicable, including the sensitivity studies used as a basis for the ECCS evaluation =cdel.
2.
If a new planc (configuration and core design) is applicable to all generic studies because it is the sa=e with respect to the generic plant design and para =eters used as a basis for a referenced co=plete spectru= defined in paragraph C, and no calculational changes resulting C
in more than 20 F change in PCT were =ade to the ECCS : del used for the referenced complete spectru=, then no new spectru= analyses are required.
The new plant =ay instead reference the applicable analysis.
Transition Break (T3)--that break size which is analyzed with both the LBM and S2M.
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C.
Minimum Recuirenents for a Comolete Break Spectrum Since it is expected that applicants will prefer to reference an applicable complete break spectru= previously conducted on another plant, this paragraph defines the mint um number of breaks required for an acceptable complete break spectrum analysis, assu=ing the cold leg pc=p discharge is established as the worst break location. The worst single failure and worst-cast reac:or coolant punp status (running or tripped) shall be established utilizing appropriate sensitivity studies.
These studies should show that the warst single failure has been justified cs a function of break size.
Each sensitivity study published during the develop =ent of the ECCS evaluation model shall be individually justified as re=aining applicable, or shall be repeated. Also, a proposal for partial loop operation shall be supported by identifying and analyzing the worst break si:e and location (i.e., idle loop versus opera:ing loop).
In addition, sufficient justifica: ion shall be previded to conclude that :he shape of the PCT versus 3reak Size curve would not be significantly altered by the partial loop configuration.
Unless this infor=ation is provided, plan:
Technical Specifications shall not permit operation with one or = ore idle reactor coolant pu=ps.
It =ust be demonstrated that the ccatain=ent design used for the break spectru= analysis is appropriate for the specific plant analyzed.
I:
should be ncted tha: this analysis is :o be perfor=ed wi:5 sa approved evaluation =odel waally in confor ance with the curren ECCS Acceptance Criteria.
1.
LBM--Cold Leg-Reac:or Coolant Pu=p Discharge a.
Three guillotine type breaks spanning a: least the range of Moody aultipliers between 0.6 and 1.0.
b.
One split type break equivalent in si:e to twice the pipe cross-sectional area.
c.
Two inter =ediate split type breaks.
d.
The large-break /s=all-break transition split.
I 2.
LBM--Cold Leg-Reactor Coolant Pu=p Suction Analyze the largest break size fro = par: I above.
If the analyses in part I above should indicate that the wors: cold leg break is an inter =ediate break size, then the largest break in the pu=p suction should be analyzed with an explanation of why the sa=e trend would not apply.
3.
LBM--Hot Leg Piping Analy:e the largest rupture in the hot leg piping.
s s
e A.
I 4.
SBM--Splits i
Analyze five dif ferent small break sizes.
One of these breaks must include the transition split break. The CFT line break =ust be analyzed for B&W plants.
This break =ay also be one of the five small breaks.
III. BOILING WATER REACTORS The generic model developed by General Electric for BWRs proposed that split and guillotine type breaks are equivalent in determining blowdown phenomena.
The staf f concluded this was acceptable and that the break area =ay be considered at the vessei no::le with a :ero loss coefficient using a two phase critical flow model.
Changes in the break area are equivalent to changes in the P.cody =ultiplier.
The minimum number of breaks required for a co=clete break spectru= analysis, assuming a suction side recirculation line break is the design basis accident (DBA) and the worst single failure has been established utilizing appropriate sensitivity studies, are shcwn in paragraph C below.
Also, a proposal for partial loop operation shall be supported by identifying and analyzing the worst break size and location (i.e., idle loop versus operating loop).
In addition, sufficient justification shall be provided to conclude that the shape of the PCT versus Break Size curve would not be significantly altered by the partial loop configuration. Unless this information is provided, plant Technical Specifications shall not permit operation with one or = ore idle reactor coolant pu=ps.
A.
BWR2, BWR3. and BWR4 Resnalvsis (Plants for which Safety Orders were issued)
If the referenced lead plant analysis is in accordance with Section III, paragraph C below, the following =inimu= number of break sizes should be reanaly:ed.
It is to be noted that the lead plant analysis is to be perfor:cd with an approved evaluation =odel wholly in confor ance with the current ECCS Acceptance Criteria. A plant =ay reference a break spectru= analysis conducted en another plant if it is the sa=e confiauration and core design.
~
Each sensitivity study published during the develop =ent of the ECCS evaluation =odel shall be individually justified as re=aining applicable, or shall be repeated.
1.
If the largest break results in the hiehest PCT:
s.
Reanaly e the li=iting break with the appropriate referenced single failure.
b.
Reanalyse the worst s=all break with the appropriate referenced j
single failure, c.
Reanalyze the transition break with the single failure and model that predicts the, highest PCT.
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o 2.
If the largest break does not result in the__hjghest PCT:
Reanalyze the li=iting break, the largest break, and a s= aller break, a.
If calculational changes have been =ade to the S3M to =ake it wholly in t'
confor=ance with 10CFR50. Appendix K,.reanaly:e the s=all break (S3M) in accordance with Section IIIC.
l 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 sa=e general design as the plant used as a basis for the lead plant analysis, but operating para =eters have changed which would increase PCT or =ctal-water reac: ion, or approved calculational changes have been made to the ECCS :odel resul:ing in more than 200F change in PCT, :he analyses of Section III, paragraph A above should be provided plus a =ini=u= of three s=all breaks (S3M),
one of which is the transition breat.
.ne snape et :he break spectru=
in the lead plant analysis should be justified as re=aining applicable, including the sensi:ivity studies used as a basis for the ECCS evaluation =odel.
2.
If a new plant (configuration or core design) is applicable to all generic studies because it is the same with respect to the generic plant design and para ecers used as a basis for a referenced co=plete spectru= defined in paragraph C, and no calcula:ional changes resulting in = ore than 200? change in ?CT were made to the ICCS =edel used for the referenced co=plete spec:ru=, then no new spectru= analyses are required.
The new plant =ay instead reference the applicable analysis.
C.
Mini =u= Recuire=ents for a Co=nlete Break Spectru=
This paragraph defines the nininu= nu=ber of breaks required for an acceptable complete spectru= analysis.
This co=plete spectrue analysis is required for each of the lead plants of a given class (SWR 2, SWR 3, SWR;,
BWRS, and SWR 6).
Each sensi:ivity study published during the develop =en:
of the ECCS evaluation model shall be individually justified as re=aining applicable, or shall be repeated.
1.
Four recirculation Line breaks at the worst location (pu=p suction or discharge), using the LBM, covering the range fro = the eransition break (T3) to the D3A, including CD coefficients of fro = 0.6 to 1.0 times the DBA.
2.
Five recirculation line breaks, us'ing the S3M, covering the range from the s=allest line break to the T3.
3.
The following break locations assu=ing the worst single failure:
a.
largest stea=line break b.
largest feedwater line bqeak
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m largest core spray line break c.
d.
largest. recirculation pump discharge or suction break (opposite sid.e of worst location)
D.
BWR4 with " Modified" ECCS Same as Section IIIC.
E.
BWR5 Same as Section IIIC.
F.
BWR6 i
Same as Section IIIC.
[
IV.
LOCA PARRETERS OF INTEREST A.
On each plant and for each break analyted, the following para =eters (versus ti=e unles: otherwise noted) should be provided on engineering graph paper of a quality to facilitate calculaticas.
--Peak clad te perature (ruptured and unruptured node)
~
--Reactor vessel pressure
--Vessel and downcocer water level (PWR only)
--Water level inside the shroud (BWR only) 1 t
--Thermal power I
--Containment pressure (PUR only) i B.
For the worst break analyzed, the following additional para =e,ters i
(versus time unless otherwise noted) should be provided on engineering l
graph paper of a quality to facilitate calculations.
The worst single failure and worst-case reactor coolant pung status will have been l
establish *ed utilizing appropriate sensitivity studies.
t
--Flooding rate (PWR only) i
--Core flow (inlet and outlet)
--Core inlet enthalpy (SWR only)
--Heat transfer coefficients
-MAPLHCR versus Exposure (SWR only)
--Reactor coolant temperature (PWR only)
--Mast > released to containment (PWR only)
--Energy released to containment (PWR only) l
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-PCT versus Exposure (Bb'R only)
- Containment condensing heat transfer coefficien: (Pk1 only)
-Hot spot flow (Pkt only)
-Quality (hottest assembly) (PG only) t
-Hot pin internal pressure
-Hot spot pellet average ce=p, rature e
-Fluid te=peratt:re (hottest asse=bly) (Pk? only)
C.' A tabulation of peak clad te=perature and =etal-water reaction (local and core-wide) shall be provided across the break spectru=.
D.
Safety Analysis Reports (SARs) filed with the h?C shall identify on each plot the run date, version nu=ber, and version date of the co=puter model utili:ed for the LOG analysis.
Should differences exist in version nu=ber or version date from the =ost current code listings made availabic to the 52C staf f, then each modification shall be identified with an assessment of i= pact upon PCT an'd =etal-water reaction (local and core-wide).
E.
A tabelation of ti=es at which significant events occur shall be provided on each plant and for each break analyzed. The following events shall be included as a mini =u=:
-Enc-of-bypass (?k2 only)
-Beginaing of core recovery (FL'R only)
-Time of rupture
-Jet pu=ps uncovered (3G only)
-MCPR (BG only)
-Time of rated spray (BWR only)
-Can quench (Bb'R only)
-End-o f-blowdown
~ Plane of interest uncovery (Sk2 only) 1
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Possible grouping of plants for the purpose of performing generic as well as individual-plant break spectrum analyses, i
CURRENT COCKETED APPLICATIONS e
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BABCOCK A!10 WILCOX CATEGORY I:
177 FA w/ Lowered Loocs Arranaement Re-analysis (Safety Order Plants):
IIA Ocyggg1,2.3 These plants must resubnit at least 3 crea<s.
(iney will co Th{geMileIsland1
(
so by reference to a cogolete IIA T
break spectrum reanalysis sua-IIA Arkansas Power 1 itted generically by 55W.)
m 2563 IIA Rancho Seco j
2772 New Ots:
Three Mile Island 2 --IIB (2)l Since these olants are the same 2772 i
design as the above plant, ney Crystal River 3
--IIB (2) v cay reference the same reanalysis i
2452 of the complete spectrum acove.
Midland 1. 2
--IIB (2) J New cps:
None CATEGORY II:
177 FA w/ Raised Leoo Arran;ement New Ots:
Davis Besse 1
--IIB Complete spectrum required.
New cps Davis Besse 2, 3
--IIB Complete spectrum required.
CATEGCRY III:
205-FA Plants New OLs:
None a
O
New cps:
'Bellefonte 1, 2
-- IIB 3 Comolete soectrum' recuired.
(Plans are for all to reference Greenwood 2, 3
-- IIB
(
a complete spectrum submitted f
probably on WPPSS.)
WPPSS 1, 4
-- IIB i
s Pebble Springs 1, 2 -- IIB )
CATEGORY IV:
145-FA Plants New Ots:
None New cps:
North Anna 3, ?
-- IIB Ccmolete scectrum recuired.
(One will prc:aaiy reference the otner.)
Surry 3, 4
-- I'B f i
e l
O I-i I
i GENERAL ELECTRIC BWR-2 Oyster Creek
-- LP*
Complete scectrum recuired.
(IIIA)**
Nine Mile Pois.*,
-- Reference only required.
(IIIA)
BWR-3 Quad Cities 2
-- LP*
Complete scectrum recuired.
(IIIA)** ;
2511 Millstone
-- IIIA - 3 breaks required 2011 Monticello
-- IIIA - 3 breaks required 1670 Dresden 2, 3
--IIiA 3
2527 L
May reference LP Quad Cities 1
-- IIIA J
2511 IIIA - 3 breaks required Pilgrim 1998 BWR-4 L'ithout fix Hatch 1
-- LP*
Comolete scectr.
recuired.
(II!A)"
2436 Peach Bottom 2, 3 -- IIIA Cc: clete scectrun recuired. One 3293 may reference ne otner.
Browns Ferry 1, 2, 3 -- IIIA j 3293 IIIA Cooper L
2381 IIIA f
f3 breaks required.
Fitzpatrick 2436 l Hatch 1 may serve IIIA - 3 breaks required Duane Arnold as a rererence I658 IIIA 5
Hatch 2' (fortheothers.
2436 IIIA Brunswick 1 2436 IIIS Shoreham IIIB Fermi IIIB Newbold
- Lead Plant
- Original break scectrum not wholly in conformance with 10CFR50, Appendix K.
43 g
)
BWR-4 Wi?' fix Brunswick 2 (Lead Plant)
~ IIIA - Cocolete soectrum 2436 recuirec...
Vermont Yankee -- IIIA - 3 breaks required (Lead Plant can be 1593 referenced, if Browns Ferry
- 1, 2, & 3 i appropr. ate)
Peach Bottom
- 2, 3 i
See preceding page Fitzpatrick*
J BWR-5 Lead Plant
-- IIIE - Comolete scectrum recuired.
Nine Mile Point 2 -- IIIB Complete spectrum required.
i.aSalle 1, 2
-- IIIB
\\ (Lead Plant can be referenced
(
by other S'WR-5 plants, if Bailly
-- IIIB appropriate.)
Zimmer
-- IIIB Susquehanna 1, 2 -- IIIB,)
BWR-6 Lead Plant
-- IIIF - Comolete s:ectrum receired.
Grand Gulf
-- IIIB Black Fox
-- IIIB Barten 1, 2, 3, 4 -- IIIB Comp'ete spectrum recuired.
(Laad Perry 1, 2
-- IIIB Plant can ce referenced by cther SWR-6 piants, if appropriate.)
Clinton 1, 2
-- IIIB Douglas Point
-- IIIB Hanford 2
-- IIIB Skagit 1, 2
-- IIIB Hartsville
-- IIIB Somerset
-- IIIB River Bend Station -- IIIB Allens Creek
-- IIIB
- May or may not have the LPCI fix
- Original break spectrum not wholly in conforpance with 10CFR50, Appendix K.
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PLANT SPECIF
!i IIIA Complete spectrum required.
Oyster Cr'eek IIIA Nine Mile Point IIIB Limerick 1, 2 IIIB Hope Creek IIIA Humboldt Bay IIIA Dresden 1 IIIA Big Rock e
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COMBUSTION ENGINEERING The following list is grouped according to similarities in design.
Some of' the older, operating plants are fairly unique, as indicated, i
and don't fall conveniently into any other groups. The list is in l
approx. chronological order.
1 1.
Palisades (Unique) -- IIA 5'
2.
Ft. Calhoun (Unique) -- ITA f
3 breaks required
- 3.
Maine Yankee (Unique) -- IIA j
4.
2560 MWt Series 3 breaks required a.
Calvert Cliffs Units 1 & 2
-- IIA
)>
Complete spectrum recuired.
b.
Hi11 stone Unit 2
-- IIB (One may reference :ne otner.)
c.
St. Lucie 1
-- IIS Complete spectrum required
- d.
St. Lucie 2
-- IIB 5.
3400 MWt Series
( 3410 MWt 217 Fue' Assemblies) h
- a. Pilgrim 2 (3470 lN )
IIB l
cmplete spectrum recuired.
b.
Forked River 1
-- IIB I
(One may reference the other.)
c'.
San Onofre 2 & 3 -- IIB d.
Waterford 3
-- IIB j
6.
Arkansas Class
(
2900 MWt 177 Fuel Assemblies) a.
R,usselville 1
-- IIS Ccmplete spectrum recuired.
(One may reference :ne otner.)
b.
Blue Hills 1
-- IIB s
Maine Yankee is unique in that it has 3 steam generators, 3 hot legs and 3 cold legs. All other CE plants have 2 steam generators, 2 hot legs and 4 colo legs.
All plants shown above listed before St. Lucie 2 are of the 14x14 fuel design. i ee All plants after, and including, St. Lucie 2 are 16x16.
I im
IIB.
Complete spectrum required 7.*
System 80 Class-(CESSAR)
These plants have not all bee 7 named yet. The utility and approx.
number of plants expected are as follows:
D a.
Duke (6) b.
WUPPS (1)
I May reference cer.:plete T
spectrum, if applicable.
Arizona Power and Light (2) c.
d.
TVA (2)
)
1
~
a G
e O
O e
b
+
i l
~
Westinghouse Ope'ratino Reactors (Safetv Order Plants)*
I 3-looc 4-1000 2-loop Ginna Surry 1/2 Yankee Rowe
~
Rewaunee Turkey Pt. 3/4 IP2 Pt. Beach 1/2 H.
B". Robinson 2
- 0. C. Cock 1 Prairie Island 1/2 Zion 1/2 i
i Ooeratina License **
2-loco I
3-1c0c 4-loco Beaver Valley 1
- Trojan
- Farley 1/2
, - Salem 1/2*
)
t North Ar.na 1/2
- Diablo Canyon 1/2*
IP-3 D. C. Cook 2 McGuire 1/2 Sequoyah I/2
- 3 breaks required (I!A).
One plant may reference another if applicable.
- Complete spectre, required.
One plant may reference another if applicable (see paragraph IIS).
a e
.lb t
l.
~~
Construction Permit **
t L
3-1000' I
4-1000 2-loop Ncrth Coast.
Sharon Harris 1/4 Byron /Braidwood 1/2 Koshkonong 1/2 Catawba 1/2 Summer 1 Floating Nuclear 1/8 Jamesport 1/2 Beaver Valley 2 l
Wisconsin Utilities Seacrcok 1/2 SNUPPS 1-5 Soutn Texas 1/2 Cemancne Peak 1/2 Watts Sar 1/2 Millstone 3 Vogtle 1/2
- Complete spectrum required.
One plant c.ay refe'rence another if applicable (see paragrapn I!S).
An g
e
t
- f. -
g-
.I 8 RANCH TECHNICAL P051T103 E!C53 18 APPLICATION CF THE SI!;;LE Fall'J2E C'4!!ER*C'; TO MAirJALLY-CCNTRCLLED I
ELECTRICALLY-CPERATED VALVES 6
1 A.
BACK0RC'J:;0 Where a single failure in an electrical syste'n can result in loss of capability to perf:r.
a safety function, tne effect on plant safety cast be evaluated. This is necessary regard-less of whetner the loss of safety function is caused my a ::. ;;nent failing :: rer'er-' a requisite rechanical etion.' or ty a cc panent performing in untestrable *echu.ica! :tien.
t This positien establishes tne ac:estacility of dist:nrecting n;cr to electrical :: ::nents of a fluid system as ene ceans of ecsigning against a single failure that -i;ht cause en un-desirable c: ;cnent acti:n. These previsions are tased on inc ass :: tion t ti tr: c:-:enect is then equivalent t: a similar c:, ::nent tnat is n:t cesigacd for electrica! :;:1tien, e.g., a valve t :t can be c::ence or closec only by direct -anual c;eratics o# tt;.el.e.
j They are also based on tne assr :tien that no sir.gic failure can ::th restore t:..ma 10 the electrical syste, and ca.se ect:nical r: tion cf tr.c co ::acnts servec ey tre ele:tri:al i
8 system. The validity of tnese assu oti:ns should te verified i.No a::!ying this rsition 1,
i e
l i; S.
BRECH TE "*;IC*L 05*7'C';
l 1.
Failurcs in botn tre ' fail to functio C scnse and the "undesirabic f. :-f:-' sa se :f cor; nents in electrical syste s Of valves and otter ' uid syste c: r e.;; s :.i" be considered in designic; aqsir.st a single failure, even :ne.ign tne.2hc c. ::cer fluid systen co r: rent ay n:t te called u::n :: fsn tion in a ;t.ea saftt. : :rati:nai sequenc2.
2.
Where it is dete.ieed that failure of an ele:trical system :: ::nant can :3.se undesired dchanical otion of a valse or otner fluid syste :: ::aen: and ids cotion results in loss of inc sysic : safety function it is acec:;2:Ic. in lieu of design changes that also may Oc a::ertable, to cis:ennect
~..ec to t.e cic:tric systers of the. valve or etner fluid system c: ;enent. The plant ted nical s:::ificati:ns shoul:
inelude a list of all electrica ly-c;erated v:lves, and 't c recaira ::st tir-s of these valves, to which the recuiremnt for re.: val of electric ;:wer is arolies in Or er ::
satisfy the single fatture criterien.
3.
Electrically-ocarated valves that are classified as " active" valves i.e.. are recuired to open or close in various safety syste Orcrational sec ceces, but are ranually-controlled, sheuld te : erated frem the.atn centrol rooc. Sucn valves -:y not be l
included among those valves frc, wnien,e er is re-oved in order to eet :ne single failure criterion unless: b) electrical pc' er i:an be, restored to the valves fr:m the l
main control reca.(b) valve operation is r.ct necessary for at least un cinutes
(
following occurrence of the event requiring such o;eration, and(:) it is de-'onstrated ON 9NQMm 7A. "
~
U Vgg
~ WU
~,
e e
l
.,~
that there is reasonable assurance that all necessary o:erator actions will be per-forred within the 11 e shown to be ade uate by the analysis. The plant technical specifications should include a list of the required positions of t anually-controlled.
electrically-c:erated valves and should identify those valves to wnien the require.
sent for remval of electric power is applied in order to satisfy the single failure criterion.
s 4
When the single failure criterion is satisfied by removal of ele:trical powee fro valves described in(2) and (3) above, tnese valves snould have r6dundant positi:n f
indication in the t ain control room and the positi:n it;dication syste Snoulc. itself,
- seet the single failure criterion.
t 5.
The phrase " electrically-o;erated valves" includes coth valves c:erated dire:tly t..
aa.
electrical device (e.g., a rctor-operated valve or a solenoid-o?erated valve; a-: t":se I
valves operated indirectly by are electrical cavice (e.g., an air-o;arated val.e a ose I
air supply is controlled by an electrical solenoid val.e).
R.EfEP.ENCES C.
E l
1.
Penorardo to R. C. DeYoung and V. A. Poore from V. 5 c11o. 0:teder 1.1973.
l e
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i e
I t
a 9
1 s
l W-4 h
n-2s
O i
BRANCH TECMICAl. POSIT!0:1 CSS 6-1 1
MINIMtat CCNTA!:;MDT PRESSURE MCCEL FOR P'.'R ECCS PERTCre'/J. E E'.'A1.UAT:C:1 l
A.
BACKC10t"O Paragraph !.0.2 of Accendia K to 10 CFR Part 50 (Def.1) requircs that the contain ent a
pressure used to evaluate the ;crfon-ance capaoility cf a pressurizec water react:r (P'.:R) emergency core cooling system (ECCS) not exceed a ;ressure calculated conservatively for that purpose. It furt*'er retuires that tns calculatien include inc effects of o;erati:n cf all installed pressure - cucirg syste.ms and precesses. Therefere, the folio.<ing cranen technical ;csition has een'cevelc:ed to provide quicance in the ;erfor ance of.?ini un containment pressure analysis. The a;;roacn descriced telc.: as;11es Only to the ECCS-related centair. ent pressure evaluatien and not to the centair?ent functional ca;atility evaluation for ;ostulated design asis accidents.
8.
BRAhr.H TECF': C*t :CS!!!P:
1.
In;ut Infer atien f:e M el a.
Initial teatair. eat <:ceral Corditions The minisun c:n*ain est gas tee:crature, nininu. contain.ent ;ressure.
and naxirum humidity inat ray te enc:antered under li:.iti..g ner :21 c:n ati,-
conditions should te usec.
b.
Initial Outside Centain ent A :f emt C:-citicns A reasonaoly ice a. rient tem:ereture external to the c:ntain :nt s":;id 2: as;:.
c.
Centair-a-t Vetu e The eaxican nct free co-tair ent volu c tr:uld be used. This axi u-free f
volume should be c:ter-ined fr:- tre gress contsin ent volu e :-inus : e v :. es I
of interrial stru:tures such as wails a-f!: ors, structural steci. -: : c: i: ant, i
and piping. The individual volume calculations sn:uid reflect the ur:er:J i a. / in J
the co: ponent vclu.as.
)
2.
Active Heat Sinks _
a.
Sprav and Fan Crot ica 5_.ste-s The operation cf all engineered safety feature centain ent heat re eval syste-s operating at eazi. u:s heat rcr eval capacity: 1.c., with all contair rnt s; ray trains crerating at maxi un flow conditions and all e e-gency fan c:oler untts operating, sh uld be asse:cd. In addition, the minicum tem; erat.re of the st: red water for the spray c: cling systc0 and the cooling water su;;1ted to the fan coolers, based on tecnnical specification li :its', snculd te assumed.
O
~
6.2.1.5-3 p
b D ((
D l
g,
'% T*
j
.,.p
- p. 9
Deviations frors th] foregofng will te ace:pted 18 't can be shown that th2 worst conditi.
regarding a single active failure, str...d water temperature, and cooling water ten erature nave been selected frort the standpoint of the overall ECCS model.
b.
Contain ent Stect Minino Vit% Soitted ECCS reter j
l The spillage of succooled ECCS =ater into the centainment provides an additional i
heat sink as the sueccoled ECCS dater eixes.:itm the stea i in the c:ntain ent.
l The effect of the steam-ma:ce mixir.g snould te considered in the contain ent pressura calculations.
c.
Contain-ent Steam Mixine ':ith Vater fre i fee N1t The water resultir.g fro, ice reiting in an ice consenser c=ntair..cnt provides an additional heat sink as the sub,ccoled.'ater mixes with the stea, nile draining from the ice c:ndenser into the lo'..ar contain cnt volume. The effe:: cf :ne
' s steam-water mixing snculd be censicered in :ne containment pressure calculati:rs.
3.
Ibssive Heat Stnks a.
Identification.
The passive neat sin <s tnat srould included in the contain-a.t eval ati:n ty icantif ing t90se structuhes and c:r er.ents wit'i' model should te estabitsr.e:
/
the contain-ent that ccuid influence the pressure res;:nse. The kir.ds cf stra:-
g tures and c:e; nents that seculd te included are listed in Tabic 1.
f Data on ;assive heat sirks nave b:en co piled fec: previ::.5 revir-s and ave
(
I been used as a basis fer tre si=;1ified r odel outlined, ale.<.
7*.is.0221 is acceptable for r.ini.:.: cen:4ir an: pressure analyses for c:nstra:;ien ter it j
applic:ti:ns, and until sacn ti. e (i.e., at the ::,eratin; iicense -evic.d ;*a a ctz.plete identificaticn of available neat si.(s can Sc race. Inis 51.:lif'ad approach has al'so :cen foll: :: for c;cratir; :lants of li:e'. sees :? ::;/ c; wit-i 1
Section 50. 6 (a)(2) of 10 C~i Part 50. Fer se:n cases. and for :: ste : ie" l
permit reviews, dere a detailed listing of
- eat sinks wt:nin ine c:ntat- :n; often cannot be ;r:vided.
- r fc110.41ng ;rececare ay be used : : ci ; e :sts*.c heat sinks within the c:ntain ent:
(1) Use the surface area and thickness of :ne prieary c:ntain-t: s:t:e1 sae11 :-
steel liner and asseciated arcters arc c:ncrete, as a;;re:ria:c.
(2) Esticate tne esposed surface arca of other steel heat sinis in acc rdance with Figure 1 and assu e an average tnickness of 3/3 inch.
(3) Model the internal cenerete structures as a slab with a thicLoess of I foe:
2 and exposed surfa:e of 160 CCO ft.
i The heat sink ther cphysical proscrties that would be acceptable are she n in Table 2.
6.2.1.5-4 i
f
At the (perating licensi stag 2 applicants should providt a detailed Itst of pess'" heat sin (s. with appropriate dimensio nd propIrtics.
b.
Heat Transfer Coefficients The following conservative condensing heat transfer coefficients for heat transfer to the exposed passive neat sinks during the blewdes.n and ; cst-blowd:wn phases of the loss-of-ccolant accident should te used (See Figure 2):
(1) During the bicwdown phase, assume a linear increase in the cendcrsing heat j
transfer ccefficient from h
=8 Stu/hr-f t
- F. at t = 0, t a ;eak g3ggg,j
'value four ti=cs greater than the eaziru, calculated c:ndensing heat trans-fer coefficient at the end of clones n using the Tagani correlaticn (Ref. 2).
- o 0.62 h
= 72.5 j.;-
2 g, = ca irun heat transfer ceef ficient Stu/hr-ft
'F where h
-Q
= pricary c:olant energy. Stu V
= net free centair-ant voh.. e. f t t,
= time interval to end of blor:d:sn. sec.
(2) During the icng-tcrn :ss bic d:v.n ;*.ase cf the accident, cnaracteri:ed ty low turbulence in t.me contain. ent at-essecre, assu ? ::ndensi. ; rea: trans'er coefft:ients 1.2 tines greater than t-ose predicted y :ne Uc ica c:ta (Ref. 3) and given in Table 3.
(3) During the transiticn chase of tre accident, tet>.aen the e-d cf t t:~. c..n sad the long-term ;ost-1c. c.:n e.ase, a reescranif c: eve-vative ec : w tu; transitien in it.e ::necesing test tr:nsfer ::affici:r.: s*.:ald ::e ass.-ed (see Figure 2).
The calculated ::ndensing heat transfer ce?fficients t: sed en the a':ve et*,d should to a;;11ed = all ex:: sed :assive heat sinks, totn etal ard ::-:rc.e. 2-d for both -painted anc un;ainted surf aces.
Heat transfer tet..een adjoining raterials in :assive heat sin (s 5-:uld ' e :ssed on the assu ; tion of no resistance to heat fl s at the m.terial i*terfa:es. An example of this is the centair. ent liner to cencrete inter' ace.
C.
REFERENCES 1.
10 CFR 150.46. " Acceptance Criteria fer Emergency C:re C: ling Systers for Lig-t i'ater Nuclear Power Reacters," and 10 CFR Part 50, A;; eat.x K. "ECCS Evaluati:r. P:dels."
j 2.
T. Tagami, "Interin Rc;:rt on Safety Assesscant and Facilities Establisa ent reciect in Japan for Peried Ending June 1965 (';o.1)," prepared for the National Aca:t r Testing station. Fceruary 23, 1966 (un;ublisnce work).
O 8.2.1.5-5
-%..~
t 3.
H. Uchid3. A. Oyama, and Y. Toga, 'Evaluatien of Post fr.:ident Cooling Syste-s of Light-Water Power
. tors," Proc. Third InternatiCnal CCnfw ace Cn the Peaceful Uses of l-Atomic Energy Volute 13. Session 3.9. United Nations. Geneva (1964).
1 i
9 I
B i
1 ii ii 4
1 h
)
MW F
I
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TA8tE 1 O.
IDENTIFICATICNOFC0!;TAl';MENTHEATS!!g 1.
Containnent Building (e.g., linte plate and external concrete walls, ficer. and sump, and liner anchors).
l 1
Contair.-ent Internal Structures (e.g.
interr.s1 separation walls and ficers, refueling 2.
f pool and f el transfer pit walls, and shielding malis).
w I
3.
Supports (e.g.. reactor vessel steam generator pu ;s. tanks...-ajor co ; r.Ents, pipe sg1 ports, and storage racks),
i 4
Uninsulated Systems and Cococnents (e.g.. cold water syste-s heating, ventilation and f
air conditionin; systees. Guns. motors. fan coolers. recs.-bir.ers, and tanks).
5.
P.iscellaneous Equipment (e.g.. ladders, gratings, cicctrical cable trays, ar.d cra es).
4 I
s
\\
{0 6.2.1. 5--7' O
.g 1
t
,=
=. _.. _..... _
~.
TABLE 2 l
HEAT 5?M THE2MOPHYSICl'. PP07ERT!ES Specific.
Ther-al Densi3y Heat Conductivity Materfa!
lb/ft Stv/lb.*F Stu/hr.ft.*F Conente 145 0.156 0.92
'I 490 0.12 27.0 I
S 9
3 l
l I
b.
s 1
e
- 6. 2.1. 5-8 vyf i.
ogm o o add u
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1I
-~.__... _
t -_
-- M
-.-a.-
.mo wm I
l
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TABLE 3 UCM!DA HEAT Tiar.5FE:t COErrICIENTS Mass Heat Transfer Mass Heat Transfer Ratio Coefficignt Ratio Coefficiget (Ib air /lb steam [
(Btu /hr.ft 'r)
(1bair/lbsteed (Stu/nr-ft' ' Q h
M 2
3 M
20 8
2.3 37 18 9
1.8 46 14 10 1.3 63 10 14 0.8 98 7
17 0.5 140 5
21 0.1 280 4
24 Y
f.
8 i
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i 6
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O i'
o
l Figure 1 Area of Steel IIcat Sinks Inside Containinent 5-
}
ee N
4 y
.S m
4J N es u o w 3
m c
.-e o
N v a H
'm m
~
p4 2
O m
62 c.
a 1
1 1
l
- 1 2
3 4
Containment Free Volume, x 10 ' f t 1
Revised 12/74 I.
O O.'
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~
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Figure 2 ll#
Condensing IIent Transfer Coefficients for St<itic Ileat Sinks i
k C3 o
w
~
l
=
o h
- 4. x h s
.w u
max Tagami 8
P Sy linear i
w I
.025(t-t )
y.
stag) e p
h=h
+ (h
-h y'd:
'i" f
i r-t ag max i
l 4
l 2
1 i
0 tb I
h
= 1.2 x h c
stag Uchida I
t
[
I o
{
e l
f:
h =8.
(j n
i I
t
)
p Time i
I blowdown, reflood l
j 1
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1
.