ML19308B270

Requests Info Required to Complete ECCS Reanalysis Review Per 741227 Order for Mod of License.Forwards Branch Technical Position Eicsb 18 & Branch Technical Position Csb 6-1

ML19308B270
Person / Time
Site:	Oconee
Issue date:	06/13/1975
From:	Purple R Office of Nuclear Reactor Regulation
To:	Thies A DUKE POWER CO.
Shared Package
ML19308B271	List: ML19308B270 ML19308B272
References
NUDOCS 7912180877
Download: ML19308B270 (13)

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Docket (3)- _

Docket Nos.

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NRC PDR (3)

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Local PDR and 50-237,

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Duke Power Company SShepard ATIN: Mr. Austin C. n ies VStello Senior Vice President ACRS (14)

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Mr. William L. Porter f-Duke Power Company P. O. Box 2178 422 South Church Strect Charlotte, North Carolina 28201 Mr. Troy B. Conner Connor, Hadlock 6.Knotts 1747 Pennsylvania Avenue, mi Washington, D. C.

20006 Oconce Public Library 201 South Spring Street g

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REQUIRED INFORMATION 1.

Break Spectrum and Partial Loop Operation,_

The information provided for each plant shall comply with the provisions of the attached memorandum entitled, " Minimum R'cquirements for ECCS Break Spectrum Submittals."

2.

Potential Boron Precipitation (PUR's Only)

The ECCS system in each plant should be evaluated by the applicant (or licensec) to shov that significant changes in chemical concentrations will not occur during the long term af ter a loss-of-coolant accident (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 assurc that boron precipitation would not compromise long-term core cooling capability following a LOCA. This review should consider all aspects of the specific plant design. including component qualification in the LOCA environment in addition to a detailed review of operating procedures.

The applicant should examine the vulnerability of the specific plant design to singic failurcs that would result in any significant boron precipitation.

3.

Single Failure Analysis A singic failure evaluation of the ECCS should be provided by the applicant (or licensee) for his specific plant design, as required by Appendix K to 10 CFR 50, Section 1.D.l.

In performing this evaluation, the effects of a single failurelor operator error that causes any manually controlled, electrically-operated valve to move to a position that could adversely affect the ECCS must be considered.

Therefore, if th'is consid-eration has not bet.. specifically reported in the past, the applicants upcoming submittal must address this consideration.

Include list of all of the ECCS valves that are currently required by the plant Technical Speeffications to have power disconnected, and any proposed plant modifications and chantes to the Technical Specifications that 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 EICSB IS from the U.S. Nuclear Regulatory Commission's Standard Review Plan is

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attached to provide you with guidance.

The singic failure eva)uction should include the potentini for passive failures of fluid systems during long term cooling following a LOCA as well as single failures of active components.

For PWR plants, the single failure analysis is to consider the potential boron concentra-problem as an integral part of long term cooling.

4.

Submery,cd Valves The applicant should review the specific equipment arrangement with-in his plant to determine if any valve motors within containment will become submerged following a LOCA.

The review should include all. valve motors that may become submerged, not only those in the safety injection system.

Valven in other systems may be needed to limit borde acid con-centration in the reactor vessel during long term cooling or may bc required for containment isolation.

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The applicant (or licensce) is to provide the following information, for each plant:

4 (1)

Whether or not any valve motors will be submerged following a LOCA in the plant being reviewed.

l (2)

If any valve motors will be flooded in their plant, the applicant (or licensce) is to:

(a)

Identify the valves that will be submerged.

(b) Evaluate the potential consequences of flooding of the valves for butin the short t.2rm and long term ECCS functions and j

containment isolation. The long term should consider the potential problem of excessive concentrations of boric acid in i

PWR's.

(c) Propose a interim solution while necessary modifications are being designed and impicmented.

(currently operating plants only).

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

5.

Containnent Pressure (PWR's Only)

The containment pressure used to evaluate the performance capability of j

the ECCS shall be calculated in accordance with the provisions of l

Branch Tc'chnical Position CSB 6-1, which is enclosed.

6.

Low ECCS.Reflood Rate (Westinghouse NSSS 0nly)

Plants that have a Westinghouse nuclear steam supply shall perform their ECC3 analyses utilizing the proper version of the evaluation model, as defined below:

(1) The December 25, 1974 version of the Westinghouse evaluation

model, i.e.,

the version without the modifications described in UCAP-8471 is acceptable for previously analyzed 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; conditions for which the December 25, 1974 and March 15, 1975 versions would be equivalent.

(2) The March 15, 1975 vehsionoftheWestinghouseevaluationmodel is an acceptabic model to be used for all previously analyzed l

plant!. for which the peak clad temperature turnaround was identi-fied to occur after the reflood rate decreased below 1.1 inches per eccond, and fo.r which steam. 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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f sen a s m MINIM 1hl REQUIRDIENTS FOR ECCS BREAK SPECTRtDi SUBMITTALS I.

INTRODUCTION The following outline shall be used as a guideline in the evaluation of LOCA break spectrum submittals. These guidelines have been formulated for contemporary reactor designs only and must be re-assessed when new reactor concepts are submitted.

The current ECCS Acceptance Criteria requires that ECCS cooling performance be calculated in accordance with an acceptable evaluation model and for a number of postulated loss-of-coolt.nt accidents of different sizes, locations and other properties cufficient to provide assurance that the entire spectrum of postulated loss-of-coolant accidents is covered.

In addition, the calculation is to be conducted with at.least three values of a discharge (C ) applied to the postulated break area, these values spanning coefficient D

the range from 0.6 to 1.0.

Sections IIA and IIIA define the acceptable break spectrum for most operating plants which have received Safety Orders. Sections IIB and IIIB define the break spectrum requirements for most CP ard OL case work (exceptions noted later). Sections IIC and IIIC provide an outline of the minimum requirements for an acceptable complete break spectrum.

Su'ch a complete break spectr'm u

l could be appropriately referenced by some plants. Sections IIID and IIIE provide the exceptions to certain plant types noted above.

A plant, due to reload a portion of its core will have previously submitted all or part of a break spectrum analysis (either by reference or by specific calculations). If it is the intention of the Licensee to replace expended fuel with new fuel of the same design (no mechanical design differences which could affect thermal and hydraulic performance), and if the Licensee intends to operate the reloaded core in compliance with previously approved Technical Specifications, no additional calculations are required.

If the reload core design has changed, the Licensee shall adopt either of Sections IIA or IIC, or of Sections IIIA or IIIC of this document, as appropriate to the plant type (BWR.or pWR). The criterion for establishing uhether paragraph A or C 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 been modified 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 spectrum analyses specified by Sections IIC or IIIC shall be submitted as a minimum (as appropriate to the plant type, BWR or PWR). Additional sensitivity studies may be required to assess the sensitivity of fuel changes in such areas as singic failures and reactor. coolant pump performance.

II.

PRESSURI7.5.'D WATER REACTORS A.

Operating Reactor Rennalyscs (plants for which Safety Orders were issued)

If calculational changes

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

• Calculational changes /Model changes--those ' revisions made to calculatio~nal techniques or fixed parameters used for the referenced complete spectrum.

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• • LBM--Large Break Model; SUM--Small Break Model l- !

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conf orm.ince with 10CFR50, Apperidix K, the.following minimum number of break sizca should be reanalyzed.

Each sensitivity study performed during the development o.f the ECCS cvaluation model shall be individually verified as remaining applicable, or shall be repeated.

A plant may reference a break if it is the same configuration spectrum analysis conducted on another plant and core design.

1.

If the largest break size results in the highest PCT:

Reanalyze the limiting break.

a.

b.

Rcanalyze two smtller breaks.in the large break region.

2.

If the largest brea'.< size does not result in the highest PCT:

Reanalyze the limiting break.

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a.

b.

Reanalyze a break larger and a break smaller than the limiting break.

If the limiting break is outside the range of Hoody multipliers of 0.6 to 1.0 (i.e., less than 0.6), then the limiting break plus two larger breaks must be analyzed.

If calculational changes have been made,to the SBM to make it wholly in small break conformance with 10CFR50, Appendix K, the analysis of the worst (SBM) as previously determined from paragraph C below should be repeated.

B.

New CP and OL Case Work A complete break spectrum should be provided in accordance with paragraph C below, except for the following:

If a new plant is of the same general design as the plant used as a 1.

basis for a referenced complete spectrum analysis, but operating parameters have changed which. would increase PCT or metal-water reaction, or approved calculational changes resulting in more than 20 F change in PCT have been nado to the ECCS model used for the referenced the analyses of paragraph A above should be provided complete spectrum, plus a minimum of three small breaks (SBM), one of which is the transition break.* The. shgpe of 'the break spectrum in the referenced j,

analysis should be justified as remaining applicabic, including the t

sensitivity studies used ay a basis for the ECCS cvaluation model, j

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If'a new plant (configuration and core design) is applicable to all jl generic studies because it is the same with respect to the generic

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,I plant design and parameters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 20 F change in PCT were made to the ECCS model used for the refercnced complete spectrum, then no new spectrum analyses are required.

The new plant may instead reference the applicable analysis.

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• Transition Break (TB)--that break size which is analyzed with both the 1

i-LBM and SBM.

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C.

Minimum Requirements for a Complete Break Spectrum Since it is expected that applicants will prefer to reference an applicable completc break spectrum previously conducted on another plant, this paragraph defines the mininum number of breaks required for an acceptable complete break. spectrum analysis, assuming the cold leg pump discherge is established as the worst break location.

The worst single failure mnd worst-case reactor coolant pump status (running or tripped) shall be established utilizing appropriate sensitivity studies.

These studies should show 'that the worst single faib", has been justified as a function of break size.

Each sensitivity stuoy published during the development of the ECCS evaluation model shall be individually justified as remaining applicabic, or shall be repeated.

Also, a proposal for partial loop

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i operation shall be supported by identifying and anal.yzing the vorst 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 citered by the partial loop configuration.

Unless this information is provided, plant Technical Specifications shall not permit operation with one or more idle reactor coolant pumps.

It must be demonstrated that the containment design used for the break spectrum analysis is appropriate for the specific plant analyzed.

It should be noted that this analysis is to be performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria.

1.

LBM--Cold Leg-Reactor Coolant Pump Discharge I

a.

Three guillotine type breaks spanning at least the range of Hoody multipliers between 0.6 and 1.0.

b.

One split type break equivalent in size to twice the pipe cross-sectional area.

c.

Two intermediate split type breaks.

d.

The large-break /small-break transition split.

I 2.

LBM--Cold Leg-Reactor Coolant Pump Suction 1

Analyze the largest break size from part 1 above.

If the analyses in part I above should indicate that the worst cold leg break is an intermediate break size, then the largest break in the pump suction should be analyzed with an-explanation of why the same trend would not apply.

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3.

LBM--Hot Leg Piping I

Analyze the largest rupture in the hot leg piping.

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4.

SBM--Splits Analyze five dif ferent small break sizes. One of these breaks must include the transition split break.

The CFT line break must be analyzed for B&W plants.

This break may also be one of the five small breaks.

III.

BOILING WATER REACTORS The generic model developed by Cencral Electric for BWRs proposed that split and guillotine type breaks are equivalent,in determining blowdown phenomena.

The staf f concluded this was acceptabic and that the break area may be considered at the vessel nozzle with a zero loss coefficient using a two phase critical flow model.

Changes in the break area are equivalent to changes in the Moody multiplier.

The minimum number of breaks required f'or a complete break spectrum 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 shown 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,~

suf ficient 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 onc or more idle reactor coolant pumps.

BWR2, BWR3, and BWR4 Reanalysis (Plants for which Safety Orders were issued)

A.

If the referenced lead plant analysis is in accordance with Section III, paragraph C below, the following minimum number of break sizes should be reanalyzed.

It is to be noted that the lead plant analysis is to be performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria.

A plant may reference a break spectrum analysis conducted on nother plant if it is the same configuration and core design.

Each sensitivity study -published duEing the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be r'epeated.

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1.

If the largest break results in the highest PCT:

Reanalyze the limiting break with 'the appropriate referenced a.

single failure.

b.

Reanalyze the worst small break with the appropriate referenced l

singic failure, Reanalyze the transition break with the singic failure and model c.

that predicts the highest PCT.

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2.

If the largest break does not result in the highest PCT:

Reanalyze the limiting break, the largest break, and a smaller break.

a.

If calculational changes have been made to the SBM to mais. it wholly in conformance with 10CFR50, Appendix K, reanalyze the small break (SBM) in accordance with Section IIIC.

4 B.

New CP and OL Case Work A complete break spectrum should be provided in accordance with Section III, paragraph C below, except fo,r the following:

If a new plant is of the same general design as the plant used as a 1.

basis for the lead plant analysis, but operating parameters have changed which would increase PCT or metal-water reaction, or approved calculational changes have been. vade to the ECCS model resulting in more than 200F cha'nge in PCT, the analyses of Section III, paragraph A above should be provided plus a minimum of three ; mall breaks (SBM),

one of which is the transition break.

The shape of the break spectrum in the lead plant analysis should be justified as remaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model.

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 j

plant design and parameters used as a basis for a referenced completc l

spectrum defined in paragraph C, and no calculational changes resulting in more than 200F change in PCT were made to the ECCS model used for the then no new spectrum analyses are required.

referenced complete spectrum, The new plant may instead reference the applicabic analysis.

C.

Minimum Requirements for a Complete Break Spectrum This paragraph defines the minimum number of breaks required for an acceptable complete spectrum analysis.

This complete spectrum analysis is f

required for cach of the lead plants of a given class (BWR2, BWR3, BWR4, BWRS, and BWR6).

Each sensitivity study published during the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated. l 1.

Four recirculation line breaksjat the worst location (pump suction or

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discharge), using the LBM, covbring the range from the transition break (TB) to the DBA, including Cp coef ficients of from 0.6 to 1.0 e

times the DBA.

Five recirculation line breaks, us'ing the SBM, covering the range 2.

from the smallest line break to the TB.

The following break locations assuming the worst singic failure:

3.

l largest steamline break a.

b.

largest feedwater line break i

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largest core spray line break d.

largest. recirculation pump dir. charge or suction break (opposite side of worst location)

D.

BWR4 with " Modified" ECCS Same as Section IIIC.

E.

BWR5 Same as Section IIIC.

F.

BWR6 Same as Section IIIC.

IV.

,LOCA PARAMETERS OF INTEREST A.

On cach plant and for cach break analyzed, the fol'. wing parameters (versus time unless otherwise noted) should be provided on engineering graph paper of a quality to facilitate calculations.

--Peak clad temperature (ruptured and unruptured node)

--Reactor vessel pressure

--Vessel and downcomer water level (PWR only) 4

--Water level inside the shroud (BWR only)

--Thermal power

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--Containment pressure (PWR only) 9 i

B.

For the worst. break analyzed, the following additional parameters (versus time unicss otherwise noted) should be provided on engineering graph paper of a quality to facilitate calculations.

The worst single failure and worst-case reactor coolant pump status will have been established utilizing appropriate sensitivity studies.

--Flooding rate (PWR only)

--Core flow (inlet and outlet) 1

--Core inlet enthalpy (BWR only)

--Itca t transfer coefficients

--HAPLilGR versus Exposure (BWR only)

-Reactor coolant temperature (PWR only)

--Mass relcasett to containment (PWR only)

,--Eacrgy released to containment (PWR only) d

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--PCT versus Exposure (BWR only)

--Containment condensing heat transfer coefficient (PWR only)

--Hot spot flow (PWR only)

--Quality (hottest asccmbly) (PWR only)

--Hot pin internal pressure

--Hot spot pel'let average temperature

--Fluid temperature (hottest assembly) (PWR only)

C.

A tabulation of peak clad temperature and metal-water reaction (local and core-wide) shall be provided across th2 break spectrum.

D.

Safety Analysis Reports (SARs) filed with the NRC shall identify on each plot the run date, version number, and version date of the computer model utilized for the LOCA analysis.

Should differences exist in version number or version date from the most current code listings made available to the NRC staff, then each modification shall be identified with an assessment of impact upon PCT an'd metal-water reaction (local and core-wide).

E.

A tabulation of times at which significant events occur shall be provided on cach plant and for each break analyzed.

The following events shall be included as a minimum:

--End-of-bypass (PWR only)

--Beginning of core recovery (PWR only)

--Time of rupture I

--Jet pumps uncovered (BWR only) l

--MCPR (BWR only)

--Time of rated spray (BWR only)

--Can quench (BWR only)

--End-of-blowdown

--Plane of interest uncovery (BWR only)

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BABCOCK AND WILCOX -

CATEGORY I:

177 FA w/ Lowered Loops rrangement Re-analysis (Safety Order Plants):

A Oc$$k8 These plants must resubmit at

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1 ast 3 breaks.

(They will do Three Mile Island 1 -- IIA so by reference to a complete f

break spectrum reanalysis sub-2535.

Arkansas Power 1

-- IIA mitted generically by B&W.)

2563 Rancho Seco

-- IIA j

2772 New Ols:

Three Mile Island 2 --IIB (2) i Since these plants are the same 2772 design as the above plant, they Crystal River 3

--IIB (2) may reference the same reanalysis 2452 of the complete spectrum above.

Midland 1, 2

--IIB (2) 4 1

New cps:.

None CATEGORY II:

177 FA w/ Raised Loop Arrangement New Ols:

Davis Besse 1

--IIB Complete spectrum required.

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New cps l

l Davis Besse 2, 3

--IIB Complete spectrum required.

CATEGORY III:

205-FA Plants New Ols:

None i1

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. BRANCH 1[CHNICAL P051710*1 E!CSB 18 APPLICATION OF THE SINGLE FAILURE CR11[RION 10 PANUALLY-CONTROLLED

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ELECTRICALLY-OPERA 1LD VALVES A.

_BACr.CROU'1D Where a s'ngic failure in an electrical system can result in loss of capability to perfort.

a safety function, the effect on' plant' safety. rust be evaluated. This is necessary regard-Icss of whether the loss of safety function is caused by a co ponent failing to perform a requisite mechanical rotion, or by a component performing an undesirable c.echtnical riotion.

This position establishes the acceptability of disconnecting po.<cr to electr[ cal co ponents i

of a fluid system as one ceans of designing against a single failure that might cause an un-desirable co ponent action. These provisions are based on the assurptfori thet the cogonent is then equivalent to a similar component that is not designed for electrical operatien.,

e.g., a valve that can be opened or closed only by direct canual operation of the valve.

They are also based on the.assu ption that no single '. allure can both restore po'.ser, to the electrical syste, and cause rechanical rotion of the co ponents served by the electrical The validity of these assur.ptions should be verified t: hen applying this pcsition.

system.

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B.

BRECH TEC$rtlCAL r051i10*;

1.

Failures in both the " fail to functioY scnse and the " undesirable unctio, sense of cocponents in electrical syste s of valves and other fluid system corr.cnents s'o;id be considered in designing against a single failure, even though the valve or ot'.er fluid systen co ponent cay not be called upon to fdctic.n in a given safety c?erational l

sequence.

2.

Where it is deterr.ined th3t failure of an electrical systcM co.conent ce". cause f

undesired re'chanical motion of a valve or other fluid syste:r. cc ponent ar.d inis j

rotion results in loss of the syster. safety function, it is acceptabic, in lieu of design changes that also ray be acceptable, to disconnect power to the electric systers The plant technical specifications should' of the valve or other fluid syster. cc.ponent.

include a list of all'cicctrically-operated valves, and 'the rroufred positions of these valves, to which the requircrent for removal of electric power is applied in order to satisfy the single failure criterion.

I Electrically-operated valves that are classified as " active" valves, i.e., are required 3.

to open or close in various safety syster operational sequences, but are ranually-1 I

controlled should be operated from the.ain control rooc. Such valves ncy not be e

included among those valves from which,ower is removed in order to nect the single l

failure criterion unless: b) electrical power can be restored to the valves from the t

i

. main control room.(b) valve cocration is not necessary for at least trn minutes fcilowing occurrence of the event requiring such operation, and(c) it is demonstrated

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.{.Fg that there is r:asonable assuranceithat all necessary operat r actions telli be per-i../

forred within the tire shown to be adequate by the analysis. The plant technical specifications should include a list of the required positions of manually. controlled, electrically-operated valves and should identify those valves to which the require-j ment for renoval of electric power is applied in order to satisfy the singic failure criterion.

4.

When the single failure criterion i; satisfied by removal of electrical power from valves described in(2) and D), above, these valves should have redundant position

~

indication in the ruin control' room and the position indication syste.m should.itself.

reet the single failure criterion.

5.

The phrase " electrically-operated valves" includes both valves operated directly by an g

electrical device (e.g., a r.otor-operated valve or a solenoid-operated valve) and those valves operated indirectly by an electrical device (e.g., an air-operated valve d ose air supply is controlled by an electrical solenoid valve).

C.

REFEREt;CES 1.

Pemorandum to R. C. OcYoung and V. A. P.oore from V. Stello, October 1,1973.

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