ML20037B542

Forwards Request for Addl Info Required to Complete Review of ECCS Reanalysis

ML20037B542
Person / Time
Site:	Dresden
Issue date:	06/18/1975
From:	Ziemann D Office of Nuclear Reactor Regulation
To:	Abel J COMMONWEALTH EDISON CO.
References
NUDOCS 8010170698
Download: ML20037B542 (25)

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beceP99R JUN 18 1975 ORB #2 Reading fi=tij KRGoller . ;q Dockst No. 50-10 TJCarter ~..... OELD g:: c.y;g OISE (31

_3 DLZiemann

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' Coc:aonwealth Edison Company A~lTN: Mr. J. S. Abel MRMDiggsd:2) ML=: Nuclear Licensing Administrator - VStello TNovak Boiling Water Reactors Post Office Box 767 JRBuchanan Chicago, Illinois 60690 TBAbernathy

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~: ACRS (14) t, DEisenhut [ Genticoen. = =: 'Ihe enclosure to this letter identifies information that we require si.:l== to complete our review of the ECCS reanalysis you will submit pursuant 35E= to our Determination by the Acting Director, Office of Nuclear Reactor if E Regulation dated April 3,1975. Some of the infonnation has previously x.:. E~ been identified (e.g., potential boron precipitation for PWRs, single failure analysis), but has been repeated here for completeness. .. + Sincerely, 6riginal signed W Ziemann Deun13 L

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Dennis L. Ziemann, Chief [;k Operating Reactors Branch #2 5j.? Division of Reactor Licensing gs}l@ =l

Enclosure:

5 Required Information (ECCS) iis cc w/ enclosure:

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? John W. Rowe, Esquire -Isham, Lincoln & Beale

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- Anthony Z. Roisman, Esquire . r22:55 Berlin,-Roiscan:and Kessler 1712 N Street, N. W.

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== - ~ 2'." b-E REQUIRED INFOR!ATION g.t 1. Brea1L Spectrum and Partial Loop Operation {m== The information provided for each plcnt shall comply with the provisions of the attached nemorandum entitled, " Minimum Requirements =:555j

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for ECCS Break Spectrum Subnittals." r===$ o 2. Potential Boron Precipitation (PhR's Only) c ;. ; i::LI The ECCS system in each plant should be evaluated by the applicant -7 (or licensee) to show that significant changes in chemical concentrations- .lZ will not occur during the long term af ter a loss-of-coolant accident g;g E.2.Z; (LOCA) and these potential changes have been specifically addressed by appropriate operating procedures. Accordingly, the applicant should. review the system capabilities and operating procedures to assure that =- boron precipitation would not compromise long-term core cooling capability . =r.=5 YEEE following a LOCA. This review should consider all aspects of the specific in

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plant design, including component qualification in the LOCA environment addition to a detailed review of operating procedures. The applicant 525 should examine the vulnerabilit.y of the specific plant design to single ...:. ZE failures that would result in any significant boron precipitation. ".fj_ q : y =, = . =, 3. Sin <:le Failure Analysis A single failure eval'uation of the' ECCS should be provided by the RER applicant (or licensee) for his specific plant design, as required by Appendix K to 10 CFR 50, Section I.D.l. In performing this evaluatien, the ef fcets of a sint,le f ailure or operator error that causes any manually q==r controlled, electrically-operated valve to cove to a position that could g . adversc3y affccc the ECCS must be considered. Therefore, if this censid-4 =g cration has not been specifically reported in the past, the applicants Include a list of all " =g upcoming submittal c.ust address this consideration. of the ECCS valves that are currently required by the plant Technical EZ Specifications to have power disconnected, and any proposed plant g2 modifications and chsnges to the Technica"_ Specifications that ci,ht be 3.Q required in order to protect against any loss of safety function caused =h 4 29=6 by this type of failure. A copy of Branch Technical Position EICSS IS from the U.S. Nuclear Regulatory Com=ission's Standard Review Plan is )) atta, ied to provide you with guidance. .n_.. The single failure evaluction should include the potential fo passive failures of fluid.,ystems during long term cooling following a l LOCA as well as single failures of active components. For PWR plants, {}~ the singic f ailure analysis is to consider the potential boron concentra- .;^~ problem as an integral part of long term cooling. m.;,y [5E5: 4. Submerged Valves The applicant should review the specific equipment arrangement with-j =j in his plant to determine if any valve motors within containment will

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become submerged following a LOCA. The review should include all valve j;] motors that may become subnerged, not only those in'the safety injection g Valves in other systems may be needed to limit boric acid con-system. centration in the reactor vessel during long term cooling or may be required for containment isalation.

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W-s.i E M ^ The applicant' (or licensee) is to. provide. the following information, for ',_.jp._. each plant: (1) Whether,or not any valve motors will be submerged following a LOCA in hh the plant being reviewed.

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If any valve motors will be floode,d in their plant, the applicant. (or licensee) is to: g5E!=% (a) Identify the valves that will be submerged. T?.] (b) Evaluate the potential consequences of flooding of the valves i-for both the sho'rt term and long term ECCS functions and ..:=EEi containment isolation. The long tern should consider the E..:q potential problem of excessive concentrations of boric acid in 32;s PWR's. .x:,:.::a= (c) Propose a interim solution while necessary modifications are siiG being designed and implemented. (currently operating plants only).

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(d) Propose design changes to solve the potential flooding problem.

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Containment Pressure (PWR's Only) EQ The containment pressure used to evaluate the performance capability of f.~f the ECCS shall be calculated in accordance with the provisions of f5j Branch Technical Position CSB 6-1, which is enclosed. 4M 6. Low ECCS Reflood Rate (Westinghouse NSSS Only)

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l Plants that have a Westinghouse nuclear steam rupply shall perforn their ECC3 analyses utilizing the proper version the evaluation nodel, as defined below: =5E

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(1) The Dece=ber 25, 1974 version of the Westinghouse evaluation

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nodel, i.e., the version without' the codifications described in jw;si FCAP-8471 is acceptable for previcusly analyzed plants for which the peak clad temperature turnaround was identified prior to the reflood Yate decreasing below 1.1 inches per second or for which -

.. 25 l f[.'M l the reflood rate was identified tc remain above 1.0 inch per second; conditions for.which the Eccomber 25, 1974 and March 15, .~;.g 1975 versions would be equiv'alent. i=a 5 (2) The March 15, 1975 version of the Westinghouse evaluation model is an acceptable codel to be used for all previously analyzed plant:, for which the peak clad temperature turnaround was identi- ~ Eeh: fi.:d to occur af ter the reflood rate decreased below 1.1 $nches per cecond, and for which steam cooling conditions (reflood rate

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less than 1 inch per second) exist prior to the time of peak clad lE=== te=perature turnaround. The March 15, 1975 version will be used ~ 4g:, for all future plant analyses. ~ss;;; ..... =1 es= l T.':.b

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MINIMUMREhREMENTSFORECCSBREAKSPECTRt[belUBMITTALS y p=::- (.... p _ =q: i:::.=::... I.* INTRODUCTION EE2 The following outline shall be used as a. guideline in the evaluation of LOCA f "] iggipi There guidelines have been formulated for break spectrum submittals. f@ggg contemporary reactor designs oniy and must be re-assessed when new reactor concepts are submitted. Hn w=m .==_c The current ECCS Acceptance Criteria requires that ECCS cooling performance x== +5n - be calculated in accordance with an acceptable evaluation model and for a i yssjy - number of postulated loss-of-coolant accidents of different sizes, locat ons i=5s and other properties sufficient to provide assurance that the entire soectrum In addition, the ERRE5 of postulated loss-of-coolant accider.ts is covered. iE5W least three values of a discharge calculation is to be conducted with at 55555 (Cp) applied to the postulated break area, these values spanning coefficient = 7 the range from 0.6 to 1.0. =:!E r~l Sections IIA and IIIA define the acceptable break spectrum fu-most operating fig i Sections IIB and IIIB define the plants which have received Safety Orders. =Evg break spectrum requirements for most CP and OL case work (exceptions noted

1. Es Sections IIC and IIIC provide an outline of the minimum require =ents

_J35 later). Such a complete break spectrum for an acceptable co=plete break spectrum. 3555 could be appropriately referenced by some' plants.. Sections IIID and IIIE jgg; provide the exceptions to certain plant types noted above. 5% A plant-due to reload a portion of its core will have previously submitted all 25== 4=E==: or part of a break spectru= analysis (either by reference or by specific If it is the intention of the Licensee to replace expended calculations). per;- fuel with new fuel of the same design (no cechanical design differences which Q{.? thermal and hydraulic performance), and if the Licensee intends i l egi; could affect to operate the reloaded core in compliance with previously approved Tecnn ca If the reload core sff.E Specifications, no additional calculations are required. 5:)sgg dssign has changed, the Licensee shall adopt either of Sections IIA or IIC, as appropriate to the plant IEEg or of Sections IIIA or IIIC of this document,The criterion for establishing whether parag R gg E type (BWR or PWR). ,j;j shall be satisfied will be determined on the basis of whether the Licensee can demonstrate that the shape of the PCT versus break size curve has not When ass 9 been modified as a consequence of changes to the reload core design. [@l the reload is supplied by a source other than the NSSS supplier, the break iB=e spectrum analyses specified by Sections IIC or IIIC shall be sub=itted as aAdditional sen r ;;;; type, BWR or PVR). (([ m mini =um (as appropriate to the plant studies may be required to assess the sensitivity of fuel changes in such areas 2;ff as single failures and reactor coolant pt=p perfor=ance. ems EE II. PRESSURIZED WATER REACTORS ' p =g=p e g:M Operating Reactor Reaaalvses (Plants for which Safety Orders were issued) iMET A. If calculational changes

• were made to the LBM** to make it wholly in

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• Calculational changes /Model changes--those revisions made to calculational

==,2 techniques or fixed parameters used for the referenced complete spectrum.

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• • LBM--Large Break Model; SB:1--S=all Break Model

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.- g=..g= tE=i-i:if conformance,with 10CFR50, Appendix E, the following minimum number of break ~ 6-sizes shnuld be reanalyzed. Each s,ensitivity study performed during the .llJll J.l development of the ECCS evaluation nodel shall be individually verified as = remaining applicable, or shall be repeated. A plant may reference a break

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spectrum analysis conducted on another plant if it is the same configuration'~ - ::n.4 and core design. 1. If the largest break size results in the highest,PE: h r-~~=d Reanalyze the limiting break. jza; - ~ a. b.. Reanalyze two smaller breaks.in the large break region. 2. If the largest break size does not result in the highest PCT: Resnalyze the limiting break. a. Ei35 (( b. Reanalyze a break larger end a break smaller than the limiting break. If the limiting break is outside the range of Moody J.E.:m multipliers of ').6 to 1.0 (i.e., less than 0.6), then the limiting i:35M break plus two larger breaks must be enslyzed, f:{r@jj !~C:: If calculational changes have been made to the SBM to make it C.olly in [Fr="p conformance with 10CFR50, Append %x K, the analysis of the worst small break p ~=T M (SBM) as previously determined from paragraph C below should be repeated. B. New CP and OL Case Work gp . d:. . ~ - A complete break spectrum should be provided in accordance with paragraph C _t s EE below, except for the following: (hjh E=E"5 1. If a new plant is of the same general des'ign as the plant used as a basis for a referenced ccmplete spectrum analysis, but operating 7;& is parameters have changed which would increase PCT or metal-water p +i Er reaction, or approved calculational changes resulting in more than 2CC F = change in PCT have been made to the ECCS model used for the referenced complete spectrum, the analyses of paragraph A above should be provided .i.g plus a minimum of three small breaks (SBM), one of which is the .l.;;:ll transition break.* The shape of the break spectrum in the referenced =.. ;9

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analysis should be justified as recaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model. f=E:

==; s ._.m 7==7 2. If a new plant (configuration and core design) is applicable to all hljG generic studies because it is the same with respect to the generic l plant design and parameters used as a basis for a referenced co=plete 2^is spectrum defined in paragraph C, and no calculational changes resulting 5., E= in more than 20 F change in PCT were made to the ECCS =odel used for 0 +5e. the referenced co :plete spectrum, then no new spectrum analyces are required. The new plant may instead. reference the applicable analysis. } =E-

• Transition Break (TB)--that break size which is analyzed with both the LBM and SBM.

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} i.=:=== 3 (#..: M...... '^ EE M.V C. Minimum Requirements for a Complete Break Spectrum

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s.... s Since it is expected that applicants will prefer to reference an applicable g 3 complete break spectrum previously conducted on another plant, this gg33 paragraph defines the minirum number of breaks required for an acceptable-

===l complete break spectrum analysis, assuming the cold leg pump discharge is. ,...Z established as the worst break location. The vorst single failure and ]~'

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worst-case reactor coolant pump status (running or tripped) shall be established, utilizing a.ppropriate sensitivity studies. These studies fiiasi should show that the worst single f ailure has been justified as a function - p.:=.y., of break size. Each sensitivity study published during the development j.] or the ECCS evaluation model shall be individually justified as remaining =i.2 applicable, or shall be repeated. Also, a proposal for partial loop g.g 5.1 operation shall be supported by identifying and analyzing the worst break size and location (i.e., idle loop versus operating loop). In addition, =B suf ficient justification shall be provided to conclude that the shape of the PCT versus Break Size curve would not be significantly altered by th: =df partial loop configuration. Unless this information is provided, plant TEF:; Technical Specifications shall not permit operation with one or more ";,] idle reactor coolant pumps. It must be de=onstrated that the containment design used for the break ~'E =5 spectrum anal'ysis is appropriate for the specific plant analyzed.- It should be noted that this analysis is to be performed with an approved @Es ~ evaluation model wholly in conformance with the current ECCS Acceptance 53 Criteria. 1. LEM--Cold Leg-Reactor Coolant Pump Discharge .7. E

E2 Three guillotine type breaks spanning at least the range of II i

a. Moody multipliers between 0.6 and 1.0.

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sps b. One split type break equivalent in size to twice the pipe p==' cross-sectional' area. c. Two intermediate split! type breaks. 4:g f d. The large-break /small-break transition split. 2. LBM--Cold Leg-Reactor Coolant Pump Suction Analyze the largest break size from part 1 above. If the analyses in d part 1 above should indicate that the vorst cold leg break is an ]~ intermediate break size, then the largest break in the pump suction { es should be analyzed with an explanation of why the same trend would irg.;== not apply. ..,..g aHE} 3. LBM--Hot Leg Piping I Analyze the largest rupture in the hot leg piping.

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I / . - -9. p 4. SSM--Splits One of these breaks must Analyze five different scall break sizes. include the transition split break. The CFT line break must be This break may also be one of the five f.5.z i analyzed for B&W plants. 3..;.f small breaks. i=.4 III. BOILISG VATER REACTORS _ W 1=. split n e generic model developed.by General Electric for DWRs proposed that and guillotine type breaks are equivalent,in determining blowdown phenomena, F the break area may be [ ne staf f concluded this was acceptabic and' that the vessel nozzle with a zero loss coefficient using a two considered at phase critical flow model. Changes in the break area are equivalent to p changes in the Moody multiplier. p.E E= ne minieun nu=ber of brecks required for a co:plete break spectrum analysis, assuming a suction side recirculation line break is the design basis accident [T{= (DBA) and the worst single failure has been 4 stablished utilizing appropriate Also, a proposal for sensitivity studies, are shown in paragraph C below. partial loop operation shall be supported by identifying and analyzing the worst E. In addition, break size and location (i.e., idle loop verr,us operating loop). suf ficient justification shall be provided to conclude that the shape of the b PCT versus Break Size curve would not be significantly altered by the partial f Unless this information is provided, plant Technical f loop configuration. Specifications shall not permit operation with one or more idle reactor l coolant pumps. F~ SW2, SWR 3, and BWR4 Reanalysis (Plants for which Safety Orders were issued A. il If the referenced lead plant analysis is in accordance with Section III, the following =inimum number of break sizes should be M" paragraph C below, reanalyzed. It is to be noted that the lead plant analysis is to be performed with an approved evaluation model wholly in confor=ance with ECCS Acceptance Criteria. A plant may reference a break spectrum analysis conducted en another plant if it is the sa e confizuration p the current and core design. ~ Each sensitivity study published.during the development of the ECCS evaluation model shall be individually justified as remaining applicable, it[ or shall be repeated. 1. If the largest break results in the highest PCT: ?. Reanalyze the limiting break with the appropriate referenced m..g. a. single failure, small brcak with the appropriate referenced b N b. Reanalyza the worst single failure. Reanalyze the tranc.ition break with the singic failure and model c. that predicts the highest PCT.

5 L.. r:= If the largest break does not t'esult in the igghest PCT: f ~

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= = =. hteanalyzethelimitingbreak,thelargestbreak,andasmallerbreak.

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If calculational changes have been adde to the SBM to make it wholly in confornance with 10CFR50, Appendix K, reanalyze the ssall break (SBM) in %] . m accordance vith Section IllC. m.1b B. New CP and OL Case Work

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.. W A complete break spectrum should b' provided in accordance with Section III, { paragraph C below, except for the following: 4 ....w.:=. ..gg If a new plant is of the same general design as the plant used as a -1. 7;E basis for the lead plant analysis, but operating parameters have 'i.ih changed which would increase PCT or metal-water reaction, or approved +:E= calculational changes have been made to the ECCS model resulting in

==}cf more than 20 F change in PCT, the analyses of Section III, paragraph A 0 above should be provided plus a minimum of three small breaks (SBM), jg one of which is the transition break. The shape of the break spectrum }.;7g3 di in the lead plant analysis should be justified as remaining applicable, including the sensitivity studies used as a. basis for the ECCS evaluation model. ....:.hi,,

== 2. If a new plant (configuratioa or core design) # 3 applicable to all =l=? to the generic "7E generic studies because it is the same with respect ~ plant design and paraceters used as a basis fvr a referenced complete spectrum defined in paragraph C, and no cales!ational changes resulting in more than 200F change in PCT were made to ine ECCS model used for the ,,ic,, referenced complete spectrum, then no new spectrum analyses are required.

== == The new plant may instead reference the applicable analysis.

==: '::::::::!b C. Minirum Recuirements for a Complete Break Spectrum .... ~. ~ 7;.,, This paragraph defines the mininun number of breaks required for an This complete spectrum analysis is e.J ? acceptable complete spectrum analysis. Z.i required for each of the lead plants of a given class (Bk'R2, Bk'R3, B'.'Rt.,

===i Bk'RS, and BWR6). Each sensitivity study published during the development

==E of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated. ~

g,. h Four recirculation line breaks at the worst location (pump suction or 1.

ra7 discharge), using the LBM, covering the range from the transition si break (T3) to the DBA, including CD coefficients of from 0.6 to 1.0 ,,, "f: f. times the DBA. ~. Five recirculation line breaks, using the SBM, covering the range 2. from the smallest line break to the TB. fE.{ rii:=i 'Ihe following break locations assuming the worst single failure: 3. =a:ii.i largest steamline break ~ T a. b. largest feedwater line break

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largest core spray line break c. .5 + 2 largg t recirculation pump discharge or suction break (opposite Ef=5 d. side of worst location)

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~.=':::. e _._.=.=.~.L.: D. BWR4 with " Modified" ECCS ish r==_= 3 Same as Section IIIC. m"pm. E.. BVR5'

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Same as Section IIIC. ..i.i..: =F s ^

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Same as Section IIIC. Ti3EE

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5.3h55b : lV. LOCA PARAMETERS OF INTEREST

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h=g= On each plant and for each break analyzed, the following parameters =~~f (versus time unless otherwise noted) should be provided on engineering [?ffil A. graph paper of a quality to facilitate calculations. L..... ~ =las:" --Peak clad temperature (roptured and unruptured node)-

4.. EE

==E= ~ g.g --Reactor vessel pressure --Vessel and downcomer water level (PWR only) I5s =i --Water icvel inside the shroud (BWR only) E=E =. =~ -Thermal power 5? . ;.. = --Containment pressure (PWR only) j=55.:: .=;.1..A For the worst break analyzed, the following additional parameters (versus time unless otherwise'noted),should be provided on engineering g#5 B. The worst single 55 graph paper of a quality to facilitate calculations. i=is=i failure and worst-case reactor. coolant pump status will have been E5::- established utilizing appropriate sensitivity studies.

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.l. --Flooding rate (PWR only) .. s=i $, 3=.=.,. -Core flow (inlet and outlet) E.=.;. 75 ,e =..m; -Core inlet enthalpy (BWR only) 37::+v +=w --Heat transfer coefficients ~ l]?". -MAPLHCR versus Exposure (BWR only)

==r -Reactor coolant ' temperature (PWR only) (PWR only) .,.g --Ma'ss released,to containment

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==. --Energy released to containment (PWR only)

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y.~. s... .._7 5!55!:.:.. p:: ;.- --PCT versus Exposure (Bk'R only) brr : --Contafnment condensing heat transfer coefficient (PWR only) [jgj3 G3EE 5.II3 .-Hot spot flow (PWR only) [2%= f --Quality (hottest asse=bly) (PWR only) .W& --11ot pin it ;ernal pressure ' pgg pmg --11ot spot pellet average temperature Ms -Fluid temperature (hottest assembly) (PWR.only) j,i,s..,i EEE - s' s A tabulation of peak clad temperature and metal-water reaction (local

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and core-wide) shall be provided across the break spectrum. a. f.-- Safety Analysis Reports (SARs) filed with the NRC shall identify on

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each plo the run date, version number, and version date of the computer D. Should differences exist in i+'~=: model utili:cd for the LOCA analysis. l5!fE version nu-ber or version date from the most current code listings made [===> availabic to the NRC staf f, then cach modification shall be identified L := g vith an assessment of impact upon PCT and metal-water reaction (local P-7 ""-

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and corc-wide). i.I[b A tabulation of tincs at which significant events occur shall be E CE The following E. provided on cach plant and for cach break analyzed. events shall be included as a minimum: Es E -Beginning of core recovery (PWR only) .[2 --End-of-bypass (PWR only)

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- = = = --Time of rupture

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+=...s --Jet pumps uncovered (Bk'R only) 5 -MCPR ~(BWR only)

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--Time of rated spray (SWR only) --Can quench (BWR only) -= p.._. e,,=.== -End-of-blowdown - ' g.p+:.f --Plane of interest uncovery (BWR only) 7 5NS=! ii:i= ' .a

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(@ (Eb 'khkh w =-, GENERAL ELECTRIC _ =., q,57 ' :x=. :. BWR-2 Oyster Creek -- LP* Complete soectrum required. (III A)** icA. .g.= Nine Mile Point -- Reference only required. (IIIA) $==cyg [g:6 5?Hi BWR.3_ Quad Cities 2 -- LP* Complete spectrum recuired. (IIIA)** 2511 .:=: I?5h:; -..... : u -- IIIA - 3 breaks required 4Es.= Millstone 2011 Monticello -- IIIA - 3 breaks required ~ iEP

== 1670 m:M Dresden 2, 3 -- IIIA 3 ' ~ f:r May reference LP ')h 2527 ..;~s Quad Cities 1 -- IIIA W.L.l... "'l$ 2511 -- IIIA - 3 breaks required Pilgrim rE Without fix Hatch 1 -- LP*- Complete soectrum recuired. (IIIA:** X5i 1993 g3 BWR _. g435

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Pe c Bottom 2, 3 -- IIIA Co nolete socctrum recuired. One sm., ""# #*#*"""C' '"' th*"* I Browns Ferry i, 2, 3 -- IIIA J

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== -- IIIA Cooper -- IIIA [' l 3 breaks required. p(.... 2381 f Fitzpatrick -- IIIA - 3 breaks required ' =4: Dua e Arnold as a reference 1658 1 Hatch 2 -- IIIA 9 / (for the others.

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i-2436 j Brunswick 1 -- IIIA [ 2436 .j;l= ---IIIB Shoreham Fermi -- IIIB =-t Newbold -- IIIB U:........ EsE;E ((,f

• Lead P1 ant
  • Original break spectrum not wholly in conformance with ICCFR50,

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~ -- itIA - Cowlete socctrum With fix Brunswick 2 (Lead Plant) 1%d requirec.** BWR-4 2436

===s. s. iermontYankee -- IIIA - 3 breaks required (Lead Plant can be

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. referenced, if 1593 appropriate) 1EQ Browns Ferry

• I, 2, & 3 3

sE=is _. u.= Peach Bottom

• 2, 3 4

See preceding page ymm1 g33.,.g ,l ..f Fitzpatrick* " -- IIIE - Cor.olete soectrum recuired. j m ::=j I BWR-5_ Lead Plant _ )

.225 Nine Mils Point 2 -- IIIB

=sEE Complete spectrum required. ?f?? \\ (Lead Plant can be referenced J.J.j LaSalle 1, 2 -- IIIB f by other BWR-5 plants,-if p :--d -- IIIB l appropriate.)

g. =-..:d Bailly f

.gg]

-- IIIB IT.'.2 Zimer GET-Susquehanna 1, 2 --IIIB) IE=CG . l$=; ; --IIIF-Cogoletesoectrumreauired. [:.";55; BWR-6_ Lead Piant } (~;-l Grand Gulf -- III B . p._.,_ x q.--... Black Fox -- IIIB I p==:: Barton 1, 2, 3, 4 - III.B Complete spectrum reauired. (Lead p=s=: F= Plant can be referenced by other =- Perry 1, 2 -- IIIB BWP.-6 plants, if appropriate.) lF!!M !X.="_F Clinton 1, 2 -- IIIB IF:Gk 12:ME:~ -- IIIB iI;.a; = Douglas Point f:h,h Hanford 2 -- IIIB a=.+.; M2 -- IIIB .:==m Skagit I, 2 ?* Hartsville -- IIIB 2.24 ~ jgruz t -- IIIB Somerset River Bend Station -- IIIB "~5=5q f.Ess Allens Creek -- IIIB 555:E l ~~"

• May or may not have the LPCI fix

FR50, 425 =

• • Original break specteum not wholly in conformance with 10CQ Appendix K.

. = = = Tl.'. ~.'.=}

= :.~2

t =l. 6 t. ('i.:.:::.. h_.',5?;- $.555..:.::IN s g l PLANT SPECIFIC [jE=M IIIA Complete spectrus required. [g.g5g Oyster Creek. g.;;;..; a .: ;s=sr IIIA , ;......g Nine Mile Point

~.:

= a 1:ww:m IIIB Limerick 1, 2 y--y .a-E=ms. IIIB

"

Hope Creek ._...= = =. IIIA e= = = Humboldt Bay h-a t.--... IIIA

== 4. Dresden 1 ,...:..r =. m III A i= E : Big Rock q=E== .:!*..5i ~ ~ " " 7.==. ...s55! nri-~ :

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• i

..q G... T... t.. I... !...EE e* g t c...2...... t-4 BRANCH TECH!i! CAL POSIT!C*l ElCSB 18 APPLICATIp'i 0F THE $1?;0'.E TAlt'.'at CRITER10* 1D FA!.'JALLY-CC!4 TROLLED 1 ELECWICALLY-CPE?ATED VALVES =g E. l.4

===N A. BACGC"*O 1 eel W ere a single failure in an ettetrical syster can result in loss of capability to perfort. This is necessary regard-e safety function, the ef fect or plant safety. tust be evaluated. 3 E~$ less of whether the loss of safety function is caused by a coaponent failing to rerf:ra a requisite eechanical riotion, or t,y a co ponent perforning an ur.desircle rechanical otion. .~~ Erd

== =q lhis position establishes the acceptability of disconnecting power to electrical cor;onents [. g.... of a fluid system as cne reans of designing against a single failure that might cause an us

iil-.}

desirabic ceronent acti n. Th:se prc<isions are based on the assu*ption that the cc.poner.; W ";. f is then equi.alent to a si.ilar co-penent that is not designed for cicetrical 0;eration. E c.g., a valve that can be opened or closed only by direct ranual c;eration of the velve. iZ base' on the assu ptien that no single failure can both restore r: er to the They ere 41s2 NM ciectrical systre and cause rechanical r.otion cf t':e co. ponents served ty the electrical The validity o' thcse assu.atio.s should be verified when a:.olying this resition. [.[1.. _ ~

=

systan. =:=.~4 ] B. PRV;tH 'IC":1 cat P"Sj,T]C,*i the "faii to f:r::tief scnse and t'e " undesirable functioC mse of ~ 1. Failures in bot' s' Id cor:por.ents in electrical syste s of valves and otner fluid syster co o:ce'.;5 be considered in desi?aino egenst a single failure._even 15 ugh the valve c-r e** er

E 5.5 fluid systen co-penent : ay not te called up;a *e isnction in a given safcty cperatforel

.:l.;

seqvence. .....=.r.

.:: =

nse 29, 3 Were it is deter-ined that fail'.re of an ele:trical syste, cc-renent cer.

2. undesired rechanical rotion of a salve or etter f!uid system c: rorer.t a :d wis .z. :7.;g rotion results in loss of the syste : safcty function, it is acceptabic, in lica of l design ch!nges that also say be a:ce? table, to discennect pe< cr to the ciectric systeas l Tbc olant technical specFications should 2 f of the valve or other fluid syste9 cc penent. 5N i'nclude a list of all electricel'y-crerated valves, and'the required positions of these l valves, to which the recuiren:rt for removal of ciectric power is applied in orter to gg l . f. =.~ satisfy the single failure criterien. l .:. 3 = =! Electrically-o: crated valves that are classi'ied as " active" valves. i.e., are required 3. to open er close in various safety syste'- eperational sequences, but are r anually-controlled, should be o: crated frem the.ain control roon..Such valves r.ay not be }.g.gg included anong those valves frc, w5ich,o.<ce is remved in order to rect the single 9.j f the "". j".i failure criterion unless: 6) electrical pcwer can be restored to the valves frot main control roce.(b) valve operation is not necessary for at least tm minutes n= fq1owing occurrence of tl.c event requiring such operation, and (:) it is dem ?:$: g;isc ( .;= 7A-27 )

=: =

."......?

h55 G..' h.T. i:2E ~ iip .- ~. 5 that there is reasonable assurance that all f.ecessary operator acL{ons will he per. ' ../ i:n.iiEl:5

--- :r-- :n forted within the ti e shown to be adequate t.y th analysis. The plant tec hical specifications should include a list of the rer.utred positions of manually-controlled.

,,,,;. 7, electrically-operated valves and should identify those valves to which the recuire-- 575E ~ snent for renoval of electric power is applied in o.#.er to satisfy oc singir failure 2.Ei: criterion. nm i._..E...M

b..h.

4 1; hen the single failure criteri. ti satisfied by rernoval of electrical power from valves described in(2) ar.d D) above these valves should have redundant position i..".. '"" indicaticn in the ruin coe.t-al roon and the position indication syste., should.itself. ((. eeet the single failurt ,.rion. th$,.7',h

==:55 5.

The phrase " electrically-o;'erated valves" ircludes both Valves socrated directly by a3 35)_hh cicetrical device (e.g., a rotor-operated valve or a solenoid-operated valve) and those .'EEin valves e, aerated indirectly by an electrical device (e.g. an air operated valve whcse f=.."55 .F ;' " air supply is controlled by an electrical solenoid valve).

• E : ::=

E*il*h C. _E L F E P.I'.CE S .ss.;.2. 2 1. Mereraad:;: to R. C. DeYoung and Y. A. Poore from V. Stello. Octcber 1.1973. mE:Eg - :::::- -[ dEEU

==

n:~~---J l t"."...L.'.*.".*.*.6, . =..;;.= .: ~. L, e --- = : y..

=. :.

AE :: I =:. Rf; .kh.... g ?.5.. ?.T: k s=;. =.: r..:...:nr.:5 l*==*:=:.' t:n ::=: IE $$.i.c.. (.... r*

==:

25-mm

_". ":Z l .......m l 1A.28 h=n %::=d L g ..a.. 1

l ~ .L .1 .= :..

r

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• L=:.=

  Mi t&C.Q

=r =,

r=~ tur.n.- .,/ N.f g= -.;

===g BRANCH TECH *i! CAL P05!T10:1 CSB 6-1

=..=..:=

MIN!?T'. C0!iTA1:. MENT-PRE 55','RE FOCEL FOR P W ECC5 PERTOR"M CE E'.*A'.UATIO!1 R ....isi;d 7"~"~'l A. BAC KGR0'J'O~~- i: ::=.:.} Paragraph !.0.2 of Aoseniix r to 10 CFR Part 50 (Pef.1) requires that the containv.cnt: j pressure used to evaluate the perforrance capability of a pressurized water reacter (P.R) cecrgency core cooling syste- (ECCS) r.ot exceed a pressure calculated conservatively fer

,EE that purpose.

't further requires that t'.e calculation include ite effects of operatica of [== ^ all installed pressure-reducing syster.s and processes. Therefore, the folloaing branch [25 technical position has been developed to provide guidance in the perfor-ance of mini s-cor.tainment pressure analysis. The approach described beloe applies only to the LCC5-related contain ent pressure evaluation ind not to the contaiteent functional capability [~3 fE5h evaluation for postulated design basis accidents, ? .knz.5 3 r. 3 I B. CRA?ct TTCF;* CAL posit!C': [ss=_B f r._.. I 1. JSyt !r.fei--atier. 'e-Wr 1 I In8tial Contain ? tit Internal Corditions ~ ~l a. 1 .l The minirvi contain e.t gas tc.serature, rininum cor.tair. ent cressure. EQ .f and riaxir.r humidity that r.ay te encountered under liridr.g nornal cetratir.'; g-i conditions shiuld 1c used. {. ..ng.;r b. Initiil Cutside Cetaie. ent t-oient Co-di icns

1. 1. 5 c=:.n.;

j 3 A rtasc-ahiy les t.C.sie... terperature external to the contairrer.t 5"c;1d be ustd. 5 5. Contf,i ent Velu*y, l;-ll[-] I c. f The raxir.:r rct frre r.o stair-ee.t volurc should be used. This raxi u-free 5.;.g. volute should be dcter ined fre : the g.est cer.uir-ent volu e nir.us Pe vcives .Q of internal structures such as walls and floors, structural stect. esfer c:ai:-ent. 5:"' g i and piping. The individual volume calculations should reflect the ur. certainty in ggg... 4?iEE f the component volu es. 3 } =;; 2. Active Heat Sir.ts .:=. a. Sprey and Fan Coolien Systems m:M The operatien cf all er.gineered safety feature contaireent heat re-oval systees .._.h - ~~f

.

y operating at maxieum heat removal capacity; i.e., with all contair. ent spray q f trains operating at maxinum flow conditions and all c:-crgency 'an cooler units g...g: operating, should be assuned. In addition. the minin c. temperature of the stored

==adi water for the spray cooling system and the cooling water supplied to the fan 4, - coolars, based on technical specification limits. should be assumed. '"..".~G {' = r 'b: m:. 9 ~. 6.2.1.5-3 "~" "1 i .:r-Q

, :==.}

O e .u

EE"mE=0 =" C":

r::.

~. - - Deviations frem the foregcing will be accepted if it can be shown that the worst , = n-i:enditions regarding a sirgic active failure, stored water teaperature, and [* '. g. cooling water temperature have been selected from the standpoint of the overall ( :=I"I ECCS r.odel. t._...~.

== isl . b. Contale ent Stea, Miring */ith $;illed rCCS Vater _.>== The spille;e of subcooled ECCS water into the containe.ent provides an additional heat sink is the sub:ooled ECCS water mixes with the steam in the contairm nt. ? The effect af the steam-weter nixing should te considered in the contain ent .j 4 pressure calculations, jgggj&

:

a ...-..j c. Contain-ent Steam Mixing t'ith Water fro Ice Pelt .. C~"-g ] The water resulting from ice relting in an ice concenser contair.s.ent provides an i. ) f, additional heat sink as ite subcooled water nixes with the steam while draining 9..~3. from the ice condenser into the lower contair. ent voluse. The cifcct Of the Mm---d i g steam-water mixing should be considered in tre containrent pressure calculatic-5. ],3 Ni-l ki 3. %ssive Feat Sinks .d[.] j JIe.n.t i.f.ip t ion a. j The passive heat sinks that should be included in the contain.ent evaluati n ] l r.odel should be established by idcntifying those structures and cc er. eats within 5% l the contain-ent t!at could in'lutece the cressure response. The kinds of struc-i tures and comp:nents that shculc be included are listed in table 1. MS y .==.:.; .( Data on rassive heat sinks have been co :iled froa previe s rcvie.es ar.d '.3.c ( been used as a t' asis fer the simplified r. ode'. outlired helu. P.is r: del is f..].{ {g' acceptable for tinira contair..cnt pressure analyses for ccnstructico rc ~it

]

ci applications, and until such tir.e (i.e., at the e;.eratir.; licers; -cvicd t*at a ..d co plete iden'tification cf avail &ble heat si.h can b2 rede. I*.is sirclified gy: ll approach has also been folle.-::: for o;.cratin; plants by lice secs :e olyi'..) wit's " E.1 r r such cases, a-d for cc ste dtio- ....s )7 Section 50.46 (a)(2) of 10 CT:t Part 53. e (:! perr.it reviews,s.here a detailed listing o' Na* sinks.:4 nin the co. tai-ent of ten cannet be ;rovided, the f:llowing precedure ray be used to r:(c1 t*e passive Z.. ~i

1 ii

3 heat sinks wit 51n the centainment

musl .nn".=4 l )4 (1) Use the surface area and thickness of the prir.ary contain.ent steel saell cc M... steel liner and associated archors and concrete. as apprceriate. 5,g (t

==

flj (2) Estinate the exposed surface arca of ot*cr steel heat sinks in accerdar.cc Eg

.

~,jl 3

with Figure 1 and assume an average thickness of 3/8 inch. . q 6 =.==v 't d. (3) Model the internal concrete structures as a slab with a thictness of I foot ..Y. _. and exposed surface of 160.000 ft. siiP" 6-5h -=: The heat sink thereophysl:a1 properties that would be acceptable are shown in ~"' f ~ 6.2.1.5-4 -c Table 2. I 3.24 e 3 2:1. J. .== iiE .} =2 m he

~ / = ib5 ' (h[E =r=2 At the operatir g license stage, applicants should' provide a detailed list sf ($d

== ::.. O passive heat sinks, with appropriate dinensions and properties.

5:i=.=

b b. 'licat Transfer Coefficieets Yb The following conservative condensing f. cat transfer ccafficients for heat transfer

g. hag.

..::.= =:: to the exposed passive heat sfr.ks tfuring the blowdown and post-blowdown phases of 1::T:J.:::. the loss-of-coolant accident should be used (See Figure ?): ~.:..y;= (1) During the bicgdoc. Phase, assume a linear increase in the conk r!ing heat transfer ceefficier.; fro-i hinitial=8 Stu/hr-f t

• F. at t = 0, to a peak

)) 7/. ?sy.:. f 'value four tizes greater than the r.axir.um calculated condensing Mat trans- [O fer coefficient at the end of blowdo.-:n using the Tagami correlaticn .I (Ref.2).*- 0.62 I"75-~" .*b [;;;;;= h = 72.0 t cax y~ = r;axirum heat transfer ceefficient. Stu/hr-ft 'T where h ~ llE55 Q = primary coolant energy. Btu E= d:5 Y = net free contain cnt volune, f1 ici-in] .g t = time interval to end of blos:dewn. sec. p bT5M (2) During the lonptern post-blowdoven phase of the accident, characterited by g5=5?f low turbulence in the con

• air. ent at osotere, assv e conc'.cnsing tea *. transfer coefficients 1.2 *ines greater then thosc-predicted by tne Uchida data ffij;f-hf

" = = = (Ref. 3) cad civen in Tabic 3. j =: ~ I end During the transition phase of the accifest. tet>:cen the ced o' ilt.uh.t. J (3) 5E= the long-ter-post-bloado.:n phase. a reasonably ccmcrvative er:o antial transition in the condensing heat trcnsfer coefficier.t should M assu~ce ' _Z_ _,..

== =: (see figure 2). bh g f The calculated condensing heat transfer coefficients based en the arcyc rethd fj should te ap;' lice to all ecosed passive teat sir.ks. toth :cetal and c: crete, and )) j

=

for both painted and unpainted surfaces. =.7.q

:::=

Heat transfer between adjoining raterials in passive heat sints should 'cc based ' .i.;..] !.n '~ E:3 on the aswmption of no resistance to heat flew at the caterial interfaces. g; exa'aple of this is the contain.cnt lincr to concrete' interface. j . M.. 2 .:...!E. :;. ..N ll C. REFERE!;C.ES 10 CFR 550.46. " Acceptance Criteria for Ercrgency Core Cooling Systers' for Light '.:ater 1 1. Ruclear Power Reactors " and 10 CFR Part 50. A;:pendix K. "ECCS Evaluation Mcdels." ?Er:5{ i T. Tagami. "Interin Report or. Safety Assessr.ents and racilitics Establish-ent Project Testing

=c2%

2. in Japan for Period Ending June 1965 (No.1)." prepared for the National Reactae Station, February 28, 1965 (unpublished work).

.==

J..a.....<

====

6.2.1.5-5 it h=.3= - ii p==.==

==

A

9

g.

.'"

::~~ T (j

n:= l vg- . := 3. H. Uchida. A. Oyama, and Y. Tota. " Evaluation of Post-Incident Cot,1ing Systes of Light- ...T.".. ~~ Water Power Rea-tors." Proc. Tttrd International Conference on the Peaceful Uses of .yllSTS Wi== U=' Atomic Energy, Volume 13. Session 3.9. United Nat, tons. Geneva (1964). l EYi... = =..... ....r~." l V:::.=:::::: in: '. u --.. t. in /. y.n..~. 1

.=::.......

1-r:r :~:=: e l.1....?..$..? '.?.,.~..

== :;d:.: .hh!.' =:- k-il ~ = ~n " -. ..; ::.:?

r:.::-

~r v u.

y e==. ( \\ f;E= =:

a:._,

l

==: g i s=.- fre j

=={ r ,' 7ii; j' 1 I

=.-

1::,, 6:- l 5::gi%i i

= -

-E.i. u,-. I / 1 'J.'... - 6.2.1.5-6 l I i 1 i e ev= .,m .m._. y,_,, . _,. j,

L*:.

e fig (p ts.si TABLE 1_ i- :r :.- a i = =. = = _e i 10tHTIFICAT101 0F C0';TAl'.v!NT HEAT SINKS _ aE..... Containment Building (e.g. liner plate and external concrete walls, floor, and sump, and 1. lineeanchors).

... j

====:: Contain ent Internal Strvetures (e.g... internal separation walls and ficors. refueling 2. g=- nool and fuel transfer pit wa'lls, and ' hic 1c'ing walls). ~ I Supports (e.g., reactor vessc1. steam generator, pa ps.. tanks, rajor co ;onents, pipe ~575 I I 3. supports. and stc. rage racks).

L.".

.::1..~..E.. Uninsulated Systems and Components (c g., cold water systems heating, ventilation, and 4. T.f i air conditioning systems, purps, notors, fan coolers, recorbiners, and tants). r== t-i='= r,iscellaneous Equip.ent (e.g., ledders, gratings, electrical cabic trays, and crancs). t E === 5. 5 $:i:$:$;5.k 2

r -:

.E2 3

E6.5.I {

===:=

'l f . = = g5

.l;:::t.

i 155 =:$2?.$ ]E~=i if"=$; l !;sss: j

u

1 1 ri=i N he q ar= li=&s?=

E:5 h5hb7:

,.. -..4, =:= '5bkkhb: a= a l =:E:: 6.2.1.5-7 j' t

=-

== ~ l

/ f t... ::= ; + se J 17: t ';l?*.? TABLE 2 HEAT $1Nt: THERMOPHY$1 CAL PROPERTIES 3.E p/ Specific Thernal Densi3y Heat Conductivity

Material,

,1b/ft Btv/15 *F Btu /hr-ft *F };... Concrete 145 0.156 0.92 Steci 490 0.12 27.0

+

== I-l..}.,

"T :.;*..

E f,:,nnh. L ~ f.i i l y, B 1 3

=;;

C [ =-- i k i il I L 8 = [t

+?

] En:h 1 1 3 1 i } 6.2.1.5-8 ( s 1 i

F .:" ~Z:

..~....

iEh f.=.i iiiiEiiiE=;

' = :lU:l z.h. Q TABLE 3

ii:.E

$~~~En UCH10A HfAT TDANSFER C0trrtCIENTS L.._-._...-= Mass Heat Transfer Mass Heat Transfer

-.A Ratio Coefficient Ratio Coefficipt j~ ~~

E" EE ~-- (1b a[ir/lb steam) (Stu/he f t' 'r) - (1bair/lbstery,1 (8tu/hr-f_t 'F1 -

==h 50 2 3 29 .. 3 20 8 2.3 37 .-=,q 9 1.8 46 -j i -18 10 1.3 63 gjjsE 14 0.8 ' 98 i..#iW.. 10 14 140 IEI"i-- 7 17 0.5 ??5'~~v :1 g";j S-21 0.1 280 4 24 g,..."... = I ..m I -- ce,--{j

.--~,..

~ ::.".' + := ..a ~. '. ' '.*.'*f.7.~.C cr.'" ? Tii:d .=.=..

cy

.....:6 .=- EM x....

a:r

. =. :.

+2== n:nn=2

n=::......

- =~=1 "~~51 =mmmmi v:. :. = = =. - 5% s .:=. 6.2.1.5-9 ...J..:.I .ei

Figurc 1 Arca of Steel IIcat Sinks Inside Containment 5-if f i. 2 i i. !.i n a be 4 .x c w m an n u O W 3 n .m a o o a N. s u x m un I e4 .a n o g 2 o Un H n h' ii24 I I 1 1 2 3 4 s Containment Free Volume, x 10k ft f i 1 l t i l-l Revised 12/74 l !i.. .., f

' ' i'i j n.,,-

.~. - mm ~ {u l': a . ! ij

. i!;.. 'i j j.

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