ML17206A935

Safety Evaluation Supporting Util Contention That Specific Tech Spec Changes Re Surveillance of High Pressure Safety Injection Throttle Valves Are Unnecessary

ML17206A935
Person / Time
Site:	Saint Lucie
Issue date:	06/19/1979
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML17206A934	List: ML17206A933 ML17206A935
References
NUDOCS 7908030264
Download: ML17206A935 (87)

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o Q UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Encl osure Safety Evaluation by the Office of Nuclear Reactor Regulation Florida Power and Light Company St. Lucie 1

Docket No. 50-335 Introduction KmLNF>7 (NOVI:I IlIll3I'7 By letter dated July 18,

1977, the NRC requested that Florida Power and Light Company (FPL) incorporate certain surveillance requirements for the St. Lucie 1 HPSI/LPSI throttle valves into the plant Technical Specifications.

Similar requests were sent to all PMR licensees.

The purpose of the surveillance requirements is to assure that total flows and flow splits to the ECCS injection points remain within specified limits.

Discussion FPL replied in a letter dated September 21, 1978 that they believe that existing requirements already provide the necessary degree of assurance and that no Technical Specification changes are needed.

The licensee stated that Technical Specification 3/4.5.2 requires that the ECCS systems at St. Lucie 1

be OPERABLE during operation in NODES 1, 2, 3 and defines certain surveillance requirements to demonstrate OPERABILITY (capability of performing its specified function).

It is FPL's interpretation that these requirements apply to the HPSI throttle valves and that appropriate surveillance and testing must be performed tq verify the settings on a routine basis and prior to declaring the ECCS subsystem OPERABLE following maintenance or system modification.

These requirements are implemented in plant procedures which are auditable by inspectors from the NRC's Office of Inspection and Enforcement.

In. addition to the plant procedures, other regulatory bases exist, including sections of 10 CFR 50, especially 50.59, and Regulatory Guide 1.33, "guali ty Assurance Program Requirements (Operation)."

Conclusion The staff has reviewed the FPL discussion and agrees that specific technical specification changes governing surveillance of HPSI throttle valves are unnecessary for the St. Lucie Unit 1 plant.

The FPL interpretation of exi sti ng Tech Specs and proceaures satisfies the intent of the NRC request.

Dated:

June 19, 1979

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4 Docket No.50-335 Florida Power 8 Light Company ATTtl:

Dr. Robert E. Uhrig Vice President Nuclear and General Engineering Post Office Box 013100 Miami, Florida 33101 During 'long tern cooling following a LOCA, the concentration of boric acid in the reactor vessel is expected to be naintained below the solubility limits by providing a flushing floi~ of coolant through the reactor vessel.

This flow is to be achieved either by simultaneous cold and hot leg injections or by sinIultaneous cold leg injection and hot leg suction.

These methods continue to be acceptable.

However, some areas of concern as to long tern acceptability of these methods have recently arisen.

In order for us to more fully evaluate these concerns we require your response to the attached requests for infomation.

DISTRIBUTION

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ST.

LUCIE UNIT HO.

1 ACRS (16)

GZech Our continuing review of the Emergency Core Cooling System (ECCS) predicted performance has led to the need for additional infomation regarding the St. Lucie Unit No.

1 plant during the long tern cooling node of operation following a postulated Loss of Accident (LOCA).

Please provide your schedule for submitting the requested information vrithin 30 days of receipt of this letter.

Sincerely, Otlyllal slgnet W

Enclosure:

Request for Additional InforE.ation P

cc w/enc a

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. Davis, Acting Chief Operating Reactors Branch F2 Division of Operating Reactors DFPICE~

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 September 27, 1977 Docket No.50-335 Florida Power 8 Light Company ATTN:

Dr. Robert E. Uhrig Vice President Nuclear and General Engineering Post Office Box 013100 Miami, Florida'3101 Gentl emen:

RE:

ST.

LUCIE UNIT NO.

1 Our continuing review of the Emergency Core Cooling System (ECCS) predicted performance has led to the need for additional information regarding the St. Lucie Unit No.

1 plant during the long term cooling mode of operation following a postulated Loss of Accident (LOCA).

During long term cooling following a LOCA, the concentration of boric acid in the reactor vessel. is expected to be maintained below the solubility limits by providing a flushing flow of coolant through the reactor vessel.

This flow is to be achieved either by simultaneous cold and hot leg injections or by simultaneous cold leg injection and hot leg suction.

These methods continue to be acceptable.

However, some areas of concern as to long term acceptability of these methods have recently arisen.

In order for us to more fully evaluate these concerns we require your response to the attached requests for information.

Please provide your schedule for submitting the requested information within 30 days of receipt of this letter.

Sincerely,.

Enclosure:

ReqI:est for Additional Information Oon K. Davis, Acting Chief Operating Reactors Branch d2 Division of Operating Reactors cc w/encl:

See next page

Florida Power 8 Light Company September 27, 1977 cc w/enclosure:

Jack R.

Newman, Esquire Lowenstein,

Newman, Reis 8 Axelrad 1025 Connecticut

Avenue, N.

W.

Washington, D.

C.

20036 Norman A. Coll, Esquire McCarthy, Steel, Hector 8 Davis 14th Floor, First National Bank Building Miami, Florida 33131 Indian River Junior College Library 3209 Virginia Avenue Ft. Pierce, Florida 33450

RE(}UEST FOR AOOITIONAL INFORtQTION In the hot leg suction method a sufficient level of coolant must be available at the bottom of the hot leg (assuming cold leg break) to prevent degraded performance because of cavitation of the residual heat removal pump.

The hot leg coolant level depends on the system pressure in the uoper plenum as determined by the total

'loop hydraulic resi stance encountered by the steam escaping from the cold )eg break.

It has been demonstrated that in most cases this resistance will be sufficiently low and the level of the water in the hot leg will be adequate.

However, for cer'tain break locations the hydraulic resistance of the steam escape path may be high enough to cause excessive loss of water level in the hot leg.

Review the ECCS for this facility and provide detailed analyses that demonstrate, regardless of the cold leg break size and location, that the hydraulic resistance of the escaping steam would be low enough to maintain hot leg water level as required to prevent pump cavi tation.

The pump used to draw water from the hot leg is designed to operate with relatively cold liquid.

Show that this pump can satisfactorily operate in the hot leg suction mode.

That is, show that the pump can simultaneously draw saturated water from the hot leg and subcooled water from the containment sump.

The procedure for hot leg suction calls for a careful control of the flow of water from the hot leg and from the containment sump.

Show that:

a.

The presently existing valve is adequate for controlling the flow.

b.

The valve is located in a sufficiently low radiation area so that it would be accessible to the operator in a post-LOCA condi tion.

c.

There is sufficient instrumentation for monitoring the flow of water from the hot leg and from the sump.

In order to assure adequate flow of water through the core during simultaneous hot and cold leg injection mode, the flow of water o

the hot and cold legs should be carefully 'balanced.

This requires the knowledge of the flow path characteristics of the system.

In view of the fact that the hot leg flow path is very complicated, involving several different lines and valves, you are requested to provide the following information:

a.

Show that all the lines in the hot leg injection flow path have sufficient cap'acity for maintaining adequate hot leg injection flow, regardless of the location of the break.

b.

Show that satisfactory procedures and instrumentation exists for monitoring hot leg injection flow.

c.

Explain in detail the procedures used for aligning the flow path for hot leg injection during the long term cooling after a LOCA.

5.

During the long term cooling mode following a postulated small

LOCA, boron precipitation is prevented by maintaining the system pressure, and therefore the saturation temperature, at sufficiently high levels.

However, the system must ultimately be depressurized and cooled in order to remove the head and inspect and/or replace the fuel.

Describe the procedures that would be used to ultimately cool down and depressurize the system following a small LOCA.

Clearly specify the equipment that would be required and show that the equipment has adequate capacity.

Oocket No.60-335 JUL 18 1977 Florida Power 8 Light Company ATTN:

Or. Robert E. Uhrig Vice President Post Office Box 013100 tifami, Florida 33101 gentlemen:

RE:

ST.

LUCIE UNIT HO.

1 A large number of PAIR High Pressure and Low Pressure Safety Injection Systems (HPSI and LPSI) utilize a common low pressure and a common high pressure header to feed the several cold (and in some cases hot) leg injection points.

Haintenance of proper flovi resistance and pressure drop in the piping system to each injection point is necessary to:

{1} prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration; (2) provide a proper flow split bebieen infection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

Many plants have either manual or motor operated valve(s) in the lines to each injection point that, have electrical or mechanical stops vrhich have been ad)usted during pre-operational testing of the plant to ensure that these flow requirements are satisified.

In view of the safety function associated with the proper setting of valves used to throttle floe in these

systems, we consider it appropriate that periodic verification be made of these valve positions.

Accordingly, ne request that you determine if throttle valves are used to obtain the required flow distribution in the HPSI or LPSI systems.

If throttle valves are used, we request that you propose changes to your technical specifications to incorporate the surveillance requirements given in the enclosure.

In the event valves are not utilized to throttle flow in your systems, you should advise us of this fact and no further action will be required.

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& Light Company

. JUL i8 1977

'Fhe above action should be completed u(ithin 60 days of receipt of

-this letter.

In the event you should desire further discussion of this matter, please contact us.

Sincerely, n

Original sig1rcd V

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Don K. Davis, cting Chief Operating Reactors Branch 82 Division of Operating Reactors

Enclosure:

Technical Specifications cc ~/enclosure:

See next page DISTRIBUTION Docket NRC PDR Local PDR ORB k'2 Reading DKDAvis RMDiggs EAReeves JWetmore OELD OI&E (3)

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Jack R.

Newman, Esquire Lowenstein,

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Washington, D.

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20036 Norman A. Coll, Esquire McCarthy, Steel, Hector 8 Davis 14th Floor, First National Bank Buildino Miami, Florida 33131 Indian River Junior College Library 3209 Virginia Avenue Ft. Pierce, Florida 33450

SAMPLE SURVEILLANCE TECHNICAL SPECIFICATIONS FOR PMR HPSI AND LPSI SYSTEM THROTTLE VALVE STOPS 1.

The correct position of each electrical and/or mechanical position stop for the following throttle valves shall be verified:

a.

Mithin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the HPSI or LPSI System is required to be operable.

b.

At least once per 18 months.

~PEI I Valve Number

~PE Valve Number l.

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3.

5.

1.

2.

3.

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5.

2.

A flow balance test shall be performed during shutdown to confirm the following minimum injection flow rates following completion of HPSI or LPSI system modifications that alter system flow characteristics:

HPSI S stem - Sin le Pum

/

Injection Leg gpm Injection Leg gpm Injection Leg gpm Injection Leg gpm Bases Injection Leg Injection Leg Injection Leg Injection Leg gPlll gPlll gpm gpm LPSI S stem - Sin le Pum The purpose of these surveillance requirements is to provide assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

Docket No. 50-335 AUG 8 6'9"6 Florida Power 8 Light Company ATTtl:

Dr. Robert E. Uhrig Vice President nuclear and General Engineering P. 0.

Box 013100 Hiami, Florida 33101 Gentlenen:

RE:.

ST.

LUCIE PLANT UHIT t)0.

1 This letter confirns a telephone conversation between the HRC and your staff on August 19, 1976.

In that conversation we stated that to satisfy the requirements for an ECCS re-evaluation in our Order for tfodification of License dated June 17, 1976 you should submit the following.

( 1)

An evaluation of. the ECCS cooling perfornance using the Combustion Engineering (CE) evaluation model with error corrections-to STRIKItl II as proposed by CE in Supple-ment 4 to CFHPD-135 dated June 1976.

These calculations must identify the worst break case.

(2)

A re-evaluation of the worst break case calculated above'~ci th a further modification to the evaluation model to preclude thy use of the nucleate boilinq heat transfer correlation during bio>>down after critical heat flux (CHF) has been predicted by the approved CHF correlation.

These analyses should be subnitte'd as soon as possible and nust be submitted before the peak linear heat generation rate liait added to your facility operating license by our Order for tiodification of License dated June 17, 1976 can be modified.

As you know, >>e are presently reviewing the naterial you submitted on July 9, 1976, to satisfy iten 1 above.

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AUG p 6 'l976 I

If you have any questions regarding the requested analyses please do not hesitate to contact us.

Sicr 1

BrPd'nh Sig ee bz:

Dennis L. Ziemann Dennis L. Ziemann, Chief Operating Reactors Branch ¹2 Division of'perating Reactors

-cc:

Jack R.

BewIman, Esquire Lowenstein, Herman, Reis 6 Axelrad 1025 Connecticut

Ayenue, H.

W.

Washington, D. C.

20036 Norman A. Coll, Esquire HcCarthy, Steel, Hector 8 Davis 14th Floor, First, National Bank Building tliami, Florida 33131 Hr. John L. ttcquigg P. 0.

Box 1408 Stuart, Florida 33494 Indian River Junior College Library 3209 Yirginia Avenue Ft. Pierce, Florida 33450 DISTRIBUTION

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Qa, Docket No. 50-335 JUL O31975 F'lorMcL Power alxl Light QcmE3any RKTAAt:

Dx. Bobert E. Uhrig Vice President of Nuclear Affairs P. 0 Box 3100 Miami, Florida 33101 Distribution:

Docket Pil NRC,PDR Local PDR LWR '1-3 File RCDeYoung RWKLecker MWilliams HRood MBirkel VHWilson ACRS (14)

EED IE Gentlemen In our June 3, 1975 letter to you, ere indicated that to cxxaplete our review of the St. Lucie Unit No. 1 HCCS re-evaluation (filed Agr33. 24, 1975), additional information is r~ed.

Specific infoxmation re-gaxding boron precipitation effects on long term ccoling, the single faQure criterion anX idle loop operation were rectuested.

Tb assure that, all the recg.axed HXS infoxmation is included in your sukmLttal, we recast that you review the attachmmt.

Pleam inform us within seven (7) days of receipt of this letter if you considec your ~ttal to be complete, or, if necessary, your schedule for sulked.ttal of additional infoxmtian. If you have any cgxeations concerning Gus matter, please feel free to contact us.

Sincerely, Oiiginal Signed b','.

D. Parr Olan D. Parr., Chief Light Niter Eleacturs Project Branch 1-3 Division of Eleactor Licensing

Enclosure:

E~uxecL Infoxnation CC'ee page 2 OPPIC93P 9URNAMC3P'L LWR 1-3 HRocd:pga 7/3/7S RL~ZlP 1-3 ODParr Porm hRC-318 (ReT. 9-53) hECM 0240 0 U 9 OOYCRNMENT PRINTINO OPPICKI I97d ddd fdd

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Xm~ein, Naenan, Reis 6 Axelrad 1025 Connecticut Avenue, N. N.

Washb@ton, D. C.

20036 Noxman A. Col1, Esq.

MCaethy, Steel, Hector ard Davis 14th F3aox, First National Hank Bui1iKrg Miami, Flor~

33131 Me. John L. Hcguigg P. 0. Bxc 1408 Stuart, Flood@

33494 OPPICCSP QORNAMCW OATCW Eorm AEC.318 (Rer. 9-33) hECM 0240 4 U, bl OOVCRNMCHT PRINTINO OPPICEI IOTA 625 IOe

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Attachment 1

REQUIRED INFORfATION Break S ectrum and Partial Loop Operation The information provided for each plant shall comply with the provisions of the attached memorandum

entitled, "Hinimum Requirements for ECCS Break Spectrum Submittals."

The ECCS system in each plant should be evaluated by the applicant (or licensee) to show that significant changes in chemical concentrations will not occur during the long term after 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 assure 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 oper'ating procedures.

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

3.

Sin le Failure Anal sis A single 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 I.D.l.

In performing this evaluation, the effects of a single failure or operator error that causes any manually controlled, electrically-operated valve to move to. a position that could adversely affect the ECCS must be considered.'herefore, if this consid-eration has not been specifically reported in the past, the'pplicants upcoming submittal must address this consideration.'nclude a list of all of the FCCS v'alves that are currently required by the plant Technical Specifications to have power disconnected, and any proposed plant modifications and changes to the Technical Specifications that might be requ'ired in order to protect against any loss of safety function caused by this type of failure.

A copy of Branch Technical Position EICSB 18 from the U.S. Nuclear Regulatory Commission's Standard Review Plan is attached to provide you with guidance.

The single failure evalu,".tion should include the potential for passive failures of fluid systems during long term cooling following a LOCA as well as single failures of active components.

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

4.

Submer ed 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.

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

4l

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

(1)

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

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

(a)

(b)

Identify the valves that will be submerged.

Evaluate the potential consequences of flooding of the valves for both the short term and long term ECCS functions and containment isolation.

The long term should consider the potential problem of excessive concentrations of boric acid in PWR's.

(c)

Propose a interim solution while necessary modifications are being designed and implemented.

(currently operating plants only).

(d)

Propose design changes to solve the potential flooding problem.

5.

Containment Pressure (Pl&'s~Onl )

Tpe containment pressure used to evaluate the performance capability of the ECCS shall be calculated in accordance with the provisions of Branch Technical Position CSB 6-1, which is enclosed.

6.

Low ECCS Reflood Rate (Westinghouse VASSS Only)

Plants that have a Westinghouse nuclear steam supply shall perform their ECCS 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 WCAP-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:farch 15, 1975 versions would be equivalent.

(2)

The l'farch 15, 1975 version of the Westinghouse evaluation model is an acceptable model to be used for alI previously analyzed plants for which the peak clad temperature turnaround was identi-fi d to occur after the reflood rate decreased below 1.1 inches per "econd, and for which steam cooling conditions (reflood rate less than 1 inch per second) exist prior to t.he time of peak clad temperature turnaround.

The ~tfarch 15, 1975 version will be used for all future plant analyses.

4 t

MINIMUM RE UIREMENTS FOR FCCS BRFAK SPECTRUM 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 A'cceptance Criteria requires that ECCS cooling performance be calculated in accordance with an acceptable evaluation model and for a number of postulated loss-of-coolant accidents of different sizes, locations and other properties sufficient to rovide assurance that the entire s ectxum of ostulated loss-of-coolant accidents is covered.

In addition, the calculation is to be conducted with at least three values of a discharge coefficient (CD) applied to the postulated break area, these values spanning the range from Oe6 to 1.0.

'ections IIA and IIIAdefine 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 and OL case work (exceptions noted later).

Sections IIC and IIIC provide an outline of the minimum requirements for an acceptable complete break spectrum.

Such a complete break spectrum cauld be appropriately reFerenced by some plants.

Sections IIXD and ZIIE 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 IIIAor IIIC of this document, as appropriate to the plant type (BWR or PWR).

The criterion for establishing whether 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'y 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 single failures and reactor coolant pump performance.

Ij.

PRESSURIZED WATER REACTORS'.

0 eratin Reactor Regnal ses (Plants for which Safety Orders were issued)

If calculational changes*

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

• <Calculational changes/Model changes those'revisions made to calculational techniques or fixed parameters used for the referenced complete spectrum.

• • LBMLarge Break Model; SBMSmall Break Model

l 0

conformance with 10CFR50, Appendix K, the following minimum number of break sizes should be reanalyzed.

Each sensitivity study performed during the development of the LCCS evaluation model shall be individually verified as remaining applicable, or shall be repe'ated.

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

). If the lar est break size results in the hi hest PCT:

a.

Reanalyze the limiting break.

b.

Reanalyze two smaller breaks, in the large break region.

2. If the lar est break size does not result in the hi hest PCT:

a.

Reanalyze the limiting break.

b.

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

If the limiting break is outside the range of Moody 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 wh'oily in

~ conformance with 10CFR50, Appendix K, the analysis of the worst s'mall break

{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.

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 made to" the ECCS model'sed for the referenced complete

spectrum, the ana'lyses of paragraph A above should 'be p'rovided plus a minimum of three small breaks'(SBM)',

one of which is the transition break.*

The shape of the break spectrum in the referenced 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 and core design) is applicable'o all generic studies because it is the same with respect to the generic plant design and parameters used as a b'asis 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 referenced complete spectrum,

'then no new spectrum analyses

-are required.

The new plant may instead reference the applicable analysis.

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

C.

Minimum Re uirements for a Com lete Break S ectrum Since it is expected that applicants will.prefer to reference an applicable complete break spectrum previously conducted on another plant, this paragraph defines the ~minim m number of breaks required for an acceptable complete break spectrum analysis, assuming the cold leg pump discharge is established as the worst break location.

The worst single failure and worst-case reactor coolant pump status (running or tripped) shall be established utilizing appropriate sensi'tivity.studies.

These studies should show that the worst single failure has been justified as a function of break size.

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

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 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'n approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria.

't 1.

LBN Cold Leg-Reactor Coolant Pump:Discharge a.

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

Dne 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'.

2.

LBM Cold Leg-Reactor Coolant Pump Suction Analyze the largest break size from part 1 above.

If the'analyses in part 1 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.

3.

LBM Hot Leg Piping Analyze the largest rupture in the hot leg piping.

4.

SBM Splits Analyze fiv'e different small break sizes.

One of these breaks must include the transition split break.

The CFT line break must be analyzed for B6W plants.

This break may al'so be one of the five small breaks.

IlI.

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 staff concluded this was acceptable 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 requirea for a

~com lets 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, 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 more idle reactor coolant pumps.

A.

J BWR2 BWR3 and BWR4 Regnal sis (Plants for which Safety Orders were issued)

$f 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'hat 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 another plant if it is the same configuration and core design.

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

l. If the lar est break results in the hi hest PCT:

a..

Reanalyze the limiting break with the appropriate referenced single failure.

b.

Reanalyze the worst small break with the appropriate referenced single failure.

c.

Reanalyze the transition break with the single failure and model that predicts the highest PCT.

~

~

2. If the lar est break does not result in the hi hest PCT:

a.

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

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

B.

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

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

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.made to the ECCS model resulting in more than 20oF change in PCT, the analyses of Section III, paragraph A

above should be provided plus a minimum of three small breaks (SBM),

one of which is the transition break.'he shape of the break spectrum in the.lead plant analysis should be )ustified as remaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model.

2.

Xf 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 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 referenced complete

spectrum,

'then no new spectrum analyses are required.

The new plant may instead reference the applicable analysis.

C.

Minimum Rc uirements for a Com lete Break S ectrum This paragraph defines the minimum number of breaks required for an acceptable complete spectrum analysis.

This complete spectrum analysis is required for each of the lead plants of a given class (BWR2, BWR3; BWR4, BWR5, and BWR6).

Each sensitivity study published during the development of the ECCS evaluation model shall be individually ')ustified as remaining'pplicable,'or shall be repeated.

1.

Four recirculation line breaks at the worst location (pump suction or discharge),

using the LBM, covering the range from the transition break (TB) to the DBA, including CD coefficients of from 0.6 to 1.0 times the DBA.

2.

Five recirculation line breaks, using the SBM, covering the range from the smallest line break'o the TB, 3.

The following break locations assuming the worst single failure:

a.

largest steamline break b.

largest feedwater line break

c'.

largest core spray line brcak d.

largest. recirculation pump discharge or suction break (opposite sige of worst location)

D.

BWR4 with "Modified" FCCS Same as Section IIIC.

F.

BVR5 Same as Section IIIC.

F.

BWR6 Same as Section IIIC.

IV.

LOCA PARAMETERS OF INTEREST A.

On each plant and for each break analyzed, the following parameters (versus time unless otherwise noted) should be provided on engineering graph paper of a quality to fa'cilitate calculations.

4 Peak clad temperature (ruptured and unruptured node)

--Reactor vessel pressure Vessel and downcomer water level (PWR only)

Water level inside the shroud (BWR only)

Thermal power Containment pressure (PWR only)

B.

For the worst break analyzed, the following additional parameters (versus time unless 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)

Core inlet enthalpy (BWR only)

Heat transfer coefficients

MAPLHGR versus Exposure (BWR only)

Reactor coolant temperature (PWR only)

Mass released to containment (PWR only)

Energy released to containment (PWR only)

gt

~

~

'w7

PCT versus Exposure

{BWR only)

-<<Containment condensing heat transfer coefficient (PWR only)

Hot spot flow (PWR only) guality (hottest assembly)

(PWR only)

->>Hot pin internal pressure Hot spot pellet average temperature Fluid temperature (hottest assembly)

(PWR only)'"

C.

A tabulation of peak clad temperature and met'al-water reaction (local and core-wide) shall be provided across the 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 and metal-water reaction (loca and core-wide).

E.

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

The following events, shall be-included as a minimum End-of-bypass (PWR -only)

't Beginning of core recovery (PWR o'nly)

Time of rupture Jet pumps uncovered (BWR only)

HCPR (BWR only) t Time of rated spray (BWR only)

Can 'quench

{BWR only)

End-of-blowdown Plane of interest uncovery (BWR only)

Possible grouping of plants for the purpose of performing generic as well as individual plant break spectrum analyses.

CURRENT DOCKETED APPLICATIONS

BABCOCK AND MILCOX.

CATEGORY I:

177 FA w Lowered Loo s Arran ement Re-anal sis Safet Order Plants Oconee 1, 2, 3

2568 Three Nile Island 1

2535 Arkansas Power 1

2563 Rancho Seco 2772 New OLs:

Three Nile Island 2

2772 Crystal River 3

2452 Midland 1, 2

New CPs:

IIA

-- IIA IIA

-- IIA

--IIB(2)

--IIB(2)

-IIB(2)

These lants must. resubmit at east 3 breaks, They will do so by reference to a complete break spectrum reanalysis sub-mitted generically by B&M.)

Since these plants are the same design as the above plant, they may reference the same reanalysis of the complete spectrum above.

None CATEGORY II:

177 FA w/Raised Loo Arran ement New OLs:

Davis Besse 1

New CPs Davis Besse 2,

3

--II8 Complete spectrum required.

Complete spectrum required.

CATEGORY III:

205-FA Plants New OLs:

None

Greenwood 2,

3

-- IIB WPPSS 1',

4 IIB Pebble Springs le 2 IIB New CPs:

Bellefonte 1, 2

-- IIB Com lete's ectrum re uired.

Plans are for a 1 to reference a complete spectrum submitted probably on MPPSS.)

CATEGORY IV:

145-FA Plants New OLs:

None New CPs:

North Anna 3, 4

Surry 3, 4

-- IIB

-- IIB Com lete s ectrum re uired.

ne will probably reference the other.)

GENERAL ELECTRIC BWR-2 Oyster Creek LP*

Com lete s ectrum re uired.

( IIIA)**

Nine Mile Point

-- Refer ence only required.

(IIIA)

BWR-3 Ruad Cities 2

2511 LP*

Com lete s ectrum re uired.

( IIIA)**

Millstone 2011 Monticello 1670 Dresden 2,

3 2527 Quad Cities 1

2511 Pil grim 1998 BWR-4 Without fix Hatch 1

2436 IIIA - 3 breaks required IIIA 3 breaks required IIIA May reference LP IIIA IIIA-3 breaks required

-- LP*

Com lete s ectrum re uired.

( IIIA)*+

Peach Bottom 3293 Browns Ferry 3293 Cooper 2381 Fitzpatrick 2436 Ouane Arnold 1658 Hatch 2

2436 Brunswick 1

2436 Shoreham Fermi Newbold 2 s 3

1,2,3 IIIA IIIA IIIA IIIA III'A -'

breaks required IIIA IIIA

-- III8 III B III B 3 breaks required.

Hatch 1 may serve as a reference for the others.

Com lete s ectrum re uired.

One may reference the other.

• Lead Plant

• • Original break spectrum not wholly in conformance with 10CFR50, Appendix K.

a

BR4 Ill ITS ik2tL dPl n IA -C~1 2436 r~euir ed. **

Vermont,Yankee

-- IIIA -

3 breaks required (Lead Plant can be 1'593 referenced, if Browns Ferry* 1, 2,

& 3 Peach Bottom* 2, 3

Fitzpatrick*

BMR-5 Lead Plant

-- IIIE -

Com Nine Nile Point 2 IIIB LaSalle 1,

2

-- IIIB Bai lly,,

-- IIIB

,Zimmer

-- IIIB Susquehanna 1,

2

-- 'IIIB appropriate)

See preceding page lete s ectrum re uired.

Complete spectrum required.

(Lead Plant can be referenced by other BWR-5 peants, if appropriate.)

BWR-6 Lead Plant IIIF -

Com lete s ectrum re uired; Cl inton 1, 2

Douglas Point Hanford 2

Skagit 1,

2 IIIB IIIB II IB II IB Har tsvi l 1 e I I IB Somerset IIIB Grand Gulf

-- IIIB Black Fox IIIB Barton 1, 2, 3, 4 --'II.B Perry 1,

2

-- IIIB Complete spectrum required.

(Lead Plant can be referenced by other BWR-6 plants, if appropriate.)

River Bend Station Aliens Creek

-- IIIB

'I I I8 May or may not have the LPCI fix

• • Original break spectrum not wholly in conformance with 10CFR50, Appendix K.

PLANT SPECIFIC Oyster Creek Nine Mile Point Lioeerick 1, 2

Hope Creek Humboldt Bay Dresden 1

Big Rock IIIA IIIA

,-- IIIB IIIB II IA'-

IIIA IIIA I

Complete spectrum required.

II II

COMBUSTION ENGINEERING The following list is grouped according to similarities in design.

Some of the older, operating plants are fairly unique, as indicated, and don't fall conveniently into any other groups.

The list is in approx. chronological order.

1.

Palisades (Unique) IIA 2.

Ft. Calhoun (Unique) -- I/A

• 3.

Maine Yankee (Unique) -- IIA 4.

2560 MWt Series 3 breaks required a.

Calvert Cliffs Units 1

5 2

-- IIA -

3 breaks required b.

Millstone Unit 2

-- IIB c.

St. Lucie 1

IIB Complete spectrum required.

{One may reference the other.)

• • d.

St. Lucie 2 IIB Complete spectrum required 5.

3400 HWt Series

(

3410 MWt 217 Fuel Assemblies)

a. Pilgrim 2 (3470 Mwt)

-.- IiB b.

Forked River 1

-- IIB c.

San Onofre 2

8 3

-- IIB d.'aterford 3

-- IIB Complete spectrum required.

(One may reference the other.)

6.

Arkansas Class

(

2900 MWt 177 Fuel Assemblies) a.

Russelville 1

b.

Blue Hills 1

IIB IIB Complete spectrum required.

{One may reference the other.)

Maine Yankee is unique in that it has 3 steam generators, 3 hot legs and 3 cold legs.

All other CE plants'ave 2 steam g'enerator s, 2 hot legs and 4 cold legs.

All plants shown above listed before St. Lucie 2 are of the 14xl4 fuel design.

All plants-after, and including, St. Lucie 2 are 16xl6.

7.

S stem 80 Class-CESSAR LIB.

Complete spectrum required These plants have not all been named yet.

The utility and approx.

1

. number of plants expected are as follows:

a.

Duke (6) b.

WUPPS (1) c.

Arizona Power and Light (2).

d.

TVA (2)

May reference complete spectrum, if applicable.

Westin house 0 e'ratin Reactors Safet Order Plants

• 2-1oo 3-1 oo 4-1 oo Ginna Kewaunee Pt.

Beach 1/2 Prairie Island 1/2 Sur ry 1/2 Turkey Pt. 3/4 8.

B'. Robinson 2

Yankee Rove IPZ D.

C.

Cook 1

Zion 1/2 0 eratin License**

2-1oo 3-1 oo 4-lop Beaver Valley 1

Farl ey 1/2

- Trojan*

- Salem 1/2*

North Arna I/2

- Diablo Canyon 1/Z IP-3

'D. C.

Cook' McGuire 1/2 Sequoyah l/2

• 3 breaks required

( IIA).

One plant may reference another if applicable.

• • Complete spectrum required..

One plant may reFerence another if applicable (see paragraph IIB).

~

~

Construction Permit **

2-1 oo 3-1 oo 4-1 oo North Coast Sharon Harris 1/4 Koshkonong 1/2 Summer 1

Beaver Valley 2 Wisconsin Utilities Byron/Braidwood 1/2 Catawba 1/2 Floating Nuclear 1/8 Jamesport 1/2 Seabr ook 1/2 SNUPPS 1-5 South Texas 1/2 Comanche Peak 1/2 Watts Bar 1/2 Millstone 3

.Vogtle 1/2

• " Complete spectrum required.

One plant may reference another if applicable (see paragraph IIB).

BRANCH TECHNICAL POSITION EICSB 18 APPLICATION OF THE SINGLE FAILURE CRITERION TO YANUALLY-CONTROLLED ELECTRICALLY-OPERATED VALVES A.

BACKGROUND Mherc a sing)e failure in an electrical systen can result in loss of capabi)ity to perfor.

a safety function, the effect on plant safety. must be evaluated.

This is necessary regard-less of whether the loss of safety function is caused by a co;puncnt failing to,"cr rm a requisite mechanical motion, or ty a coo.poncnt performing an undesirable

~echarrical notion.

This position establishes the acceptability of disconnecting Power to electrical components of a fluid system as one neans of designing against a single failure that right cause an un-desirable co..ponent action.

These provisions are based on the assumption that the co.poncr.t is then equivalent to a sinilar coi".,ponent that is not designed for electrical operation, e.g.,

a valve that can be opened or closed only by direct,.anual operatiori of tht valve.

They are also based on the assu:-ption that no single failure can both restore r c.: r to the electrical system and cause mechanical rrotion of the cor.poncnts served by 'thc electrical systen.

The validity.of these assu.-ptions should b

verified when applying this position.

~ r BRANCH TEC!!"iCAL POSITION Failures in. both the "fail to func'on stlrse ard the "undesirable functio" '

.nsc o

components in electrical systems of-valves arid other fluid systc.. co-.:::rents s-o.r be considered in. designing against a single failure," even th,ugh t,he va.Vc or'".c r fluid systcrr co-.,:poncnt may not be called upon tc function in a ci.er. sartty c,"erationar sequence.

2.*

Hhere it is deteri.;ined that fai)ure of an electrical system co-..ponent ca".

ca se undesired r:,e'chanical motion of a valve or other f)uid system cc.-.ponent ard : irs rrotion results in loss nr tl c system safety function, it is acceptab)c, iin )ieu of design changes that also may be acceptable, to disconnect pc>>er to the electric systers of the valve or other fluid system ccmponcnt.

The plant technical specifications should r,

h include a list of al) electrically-operated

valves, and tie required positions o.

these

valves, to which the requirement for removal nf electric po>>er is applied in order to satisfy the single failure criterion.

3.

Electrically-opciated valves that are classified as "active" valves, i.e., are required to open or close in various safety systcr. operational sequences.

but are -anually-controlled, should be operated from the main control room.

Such valves may not be included among those valves from which,ower is removed in order to meet the single failure criterion unless:

(~) electrical power can be restored to the valves from the main control roon,(b) valve operation is not necessary for at least ten ninutes following occurrence of the event requiring such operation, and (c) it is demonstrated 7A-27

that there is reasonable assurance that all necessary operator actions will be per-formed within the time shown to be adequate by the analysis.

The plant technical specifications should include a listof the required positions of manually-controlled, electrically-operated valves and should identify those valves to which the require-ment for removal of electric power is applied in order to satisfy the single failure criterion.'

~

~.,P 4.

hhen the single failure criterion is satisfied by removal of electrical power from valves described in (2) and (3 ), above, these valves should have redundant position indication in the main control room and the position indication syste'.

should, itself, meet the single failure criterion.

5.

The phrase "electrically-operated valves" includes both valves operated direc.tly bg an eleCtriCal deViCe (e.gme a mOtOr-Operated ValVe Or a SOlenOid-Operated ValVe) and thOSe valves operated indirectly by an electrical device (e.g.,

an air-operated valve whose air'supply is controlled by an electrical solenoid valve).

C.

REFEIIERCES 1.

Hemorar.dure to R.

C.

DcYouog.eod V. R. Iaoore from V. Stello, 0."tuber 1, 1973.

7A-28

BRANCH TECHNICAL POSITION CSB 6-1 HINIHUH CONTAINHENT PRESSURE HOOEL FOR PWR ECCS PERFORHANCE EVALUATION A.

BACKGROUNO Paragraph 1.0.2 of Appendix K to 10 CFR Part 50 (Ref.

1) requires that the containment pressure used to evaluate the performance capability of a pressurized water reactor (PWR) emergency core cooling system (ECCS) not exceed a pressure calculated conservatively for v

that purpose.

It further requires that the calculation include the effects of operation of all installed pressure-reducing systems and processes.

Therefore, the following branch technical position has been developed 4o provide guidance in the perfor. ance of minirum containment pressure analysis.

The approach described below applies only to the ECCS-related containment pressure evaluation and not to the containment functional capability evaluation for postulated desiqn basis accidents.

'Q B.

BRANCH TECHNICAL POSITION 1.

~ln ut Information for Yodel a.

Initial Contain-..ent Internal Conditions The minimum containment gas temperature, rin'imur containr.'.ent

pressure, and maximum humidity that may be encountered under limiting norr.:al operatinq conditions shnuld le used.

b.

Initial Outside Contain-..ent Ambient Conditions A reasonably low ar:bient temperature external to the contair.-.ent sf cuIC be used.

C.

Containment Volur!e The maximum net free containrent volume should be used.

This raxl,".im free volume should be deter'.ined fronn thc gross containment volut"e ainus:he volu-..es of interrial structures such as walls and floors, structural steel, njor ecuipment,g and piping, The individual volume calculations should reflect the uncertainty in the component volur,es.

2.

Active Heat Sinks a.

~S va and Fan Coulton ~S stens The operation of ail engineered safely feature eontainnent heat re.,oval systens operating at maximum heat removal capacity; i.e., with all containment spray trains operating at maximum flow conditions and all emergency fan cooler units operating, should he assumed.

In addition, the minimum temperature of the stored water for the spray cooling system and the cooling water supplied to the fan

coolers, based on technical specification limits, should be assumed, 6.2.1.5-3

Deviations om the foregoing will be accepted if it be shown that the worst conditions regarding a single active failure, stored water temperature, and cooling water temperature have been selected from the standpoint of the overall ECCS model.

h.

Cnntainment Stean Nixie

",ith S i11ed CCCS Paten The spillage of subcooled ECCS water into the containment provides an additional heat sink as the subcooled ECCS water mixes with the steam in the containment.

The effect of the steam-water nsixing should be considered in the containment pressure calculations.

c.

Containment Steam Hixin

! 1th 'Water from Ice 'Kelt The water resulting from ice melting in an ice condenser containment provides an additional heat sink as the subcooled water mixes with the steam <<hile draining from the ice condenser into the lower containment volume.

The effect of.the steam-water mixing should be considered in the containment pressure calculations.

3.

Passive Heat Sinks a.

Identification The passive heat sinks that should be included in the containment evaluation model should be established by identifying those structures and components wit> in the containment that could influence the pressure response.

The kinds of struc-tures and components that should be included are listed in Table l.

Data on passive heat sinks have been compiled from previo's revie.is and have been used as a basis for the simplified model outlined below.

This rodel is acceptable for minimum containment pressure analyses for construction pes.'lit applications, and until such time (i.eae at the operating license'eview';

tnat a

complete identification of available heat sirks can be made, Tnis simplified approach has also been followed for operating plants by licensees co->eyirg <<itn Section 50.46 (a)(2) of 10 CFR Part 50.

For such cases, and for constr;:ction permit reviews, where a detailed 'listing of heat sinks within the contain-ent often cannot be provided, the following 'procedure may be used to.-..odel

ne >ass ve heat sinks within the containmcnt:

(1)

Use the surface area and thickness of the primary containment steel shell cr steel liner and associated anchors and concrete, as'ppropriate.

(2)

Estimate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assume an average thickness of 3/8 inch.

(3)

Model the internal concrete structures as a slab with a thickness of 1 foot and exposed surface of 160,000 ft.

2 The heat sink thermophysical properties that would he acceptable are shown in Table 2.

6.2.1.5-4

~

~

At the o ating license stage, applicants should vide a detailed list,of passive heat sinks, with appropriate dimensions and properties.

b, Keat Trans(cr Coefficients The folloving conservative condensing heat transfer coefficients for heat transfer to the exposed passive heat sinks during the blowdown and post-blowdown phases of the loss-of-coolant accident should be used (See Figurc 2):

0 (1)

During the blowdown phase, assume a linear. increase in the condensing heat transfer coefficient from h 8 Btu/hr-ft - F, at t 0, to a peak 2

0 initial

'value four times greater than the maximum calculated condensirg heat trans-fer coefficient at the end of blowdown, using the Tagahi correlation (Ref. 2),

0.62 h

< 72.5 where h

max>r..um heat transfer coefficient, Btu/hr-ft -'F max

~ primary coolant energy, Btu 3

V

~ net free containment volume, ft t

time interval to end of blowdown, scc.

p h

(2)

During the long-term post-bio>>down phase of the accident, characterized Ly low turbulence in the containment atcosphere, assume condensing heat transfer coefficients 1.2 tires greater than those predicted by the Uchida data (Ref. 3) and given in Table 3.

(3)

During the transition phase of the accident, Letween thc end of blo~donn a~d the long-term post-blowdown phase, a reasonably conservative expo'en ia I transition in the condensing heat transfer coefficient should ';" ass."e" (See Figure 2).

I The calculated condensing heat transfer coefficients based on the a.cvc -e'..",;"

'hould be applied to all exposed passive heat sirks,,both metal and co, cre:e, and for both.painted and unpainted surfaces.

Keat transfer between adjoining materials in passive heat sinks should ce based on the assumption of no resistance to heat flow at the material interfaces.

An example of this is the containment liner to concrete'ntci face.

C.

REFERENCES l.

10 CFR

$ 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Mater Nuclear Power Reactors,"

and 10 CFR Part 50, Appendix K, "ECCS Evaluation vodels."

2.

T. Tagami, "Interim Rcport, on Safety Assessments and Facilities Establish.-.cnt Project in Japan for Period Ending Junc 1965 (No. 1)," prepared for the National React~~ Testing

Station, February 28, 1966 (unpublished work).

6.2.1.5-5

H. Uchida, A.'y, and Y. Toga, "Evaluation of Post-Inci t Cooling Systems of Light-Mater Po~er Reactors,"

Proc. Third International Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, United Nations, Geneva (1964).

6.2.1.5-6

y

~

~

a Q

TA8LE I 10ENTIFICATION OF CONTAINNENT NEAT SINKS Containment 8uilding (e.g., liner plate and external concrete walls, floor, and

sump, and liner anchors).

2, Containment Internal Structures (e.g., internal separation walls and floors, refueling pool and fuel transfer pit walls, and shielding walls).

3.

Supports (e.g., reactor vessel, steam generator,

pumps, tanks,. major components, pipe

supports, and storage racks).

4.

Uninsulated Systems and Components (e.g.,

cold water systems, heating, ventilat on, and air conditioning systems.

pumps, motors, fan coolers, reco..biners,,and tanks).

5.

Hiscellaneous Equipment (e.g.,

ladders, gratings, electrical, cable trays, and cranes).

6.2. I. 5-7

'LP TABLE 2 HEAT SINK THERNDPHYSICAL PROPERTIES Haterial Concrete Steel Densi)y

~th ft 145 490 Specific Heat htufth-'F

0. 156

0. 12 Thermal Conductivity

~Btu hr-ft-'F 0.92

'7.0 6.2,1.5-8 uu r

~ 8 7~ ~ %,~ frrutupuuu

~

u Ft uu u

P

. ~

TABLE 3 Hass Ratio lb air lb steaa UCHIOA HEAT TRAHSFER COEFFICIENTS Heat Transfer Hass Heat Transfer Coefficient Ratio Coefficient L'"'""'""L 50 20 18 14 10 7

5 4

2 8

9 10 14 17 21 24 2 '

1

~ 8 1.3 0.8 0.5 0.1 29 37 46 63 98 140 280 6.2. 1. 5-9

Figure 1

Area of Steel lieat Sinks Tnside Containment 2

3 Containment Free Volume, x 10't 6

3.

Revised 12/74

Figure 2

Condensing Heat Transfer Coefficients for Static Heat Sinks 4J O

W CJO max Tag ami linear I

.025(t-t) h=h

+(h

-h

) e

.p stag max stag C4 C

~ll Ol'or0 0

I t

I P

I 4 1 owdoMn I

re flood I

I I

stag Uchida h

= 1.2xh

Docket No. 50-335 JUN 0-3

]97g Distribution NRC PDR OELD-Local PDR.

~E (3)

Docket File~~ HRood LWR 1-3 File

%Wilson RCDeYoung TR Branch Chiefs FSchroeder LWR 1 Branch Chiefs AKenneke JPanzarella RWKlecker ACRS (14)

Floxida Power 6 Light Company ATTN:

Dr. Robert E. Uhrig Vice President of Nuclear Affairs P. 0. Box 3100 Miami, Florida 33101 Gentlemen:

Before we can complete our review of youx ECCS re-evaluation for St. Lucie 1, which was docketed on April 24, 1975, additional information is required.

The specific information requested is contained in the Enclosure and concerns boron precipitation effects on long term cooling, the single failure criterion, and idle loop operation.

These items have been discussed previously with your representatives.

To maintain our review schedule, we will need your response to the items in the Enclosure by July 3, 1975. If you cannot pxovide this information by this date, please inform us within seven (7) days after receipt of this letter so that we may revise our schedule.

Sincerely, Original Signed bg Olan Parr Olan D. Parr, Chief Light Water'Reactors Pro)ect Branch 1-3 Division of Reactor Licensing

Enclosure:

Request for Additional Information cc:

See page 2

OPPICC~

SURNAMC~

OATR~

LWR 1-3 IM HRoodrmck t'/gJ75 LiS 3p3 ODParr 6/S /75 Form AEC.518 (ROT. 9.$ 3) AECM 0240 0 U, 0I OOVRRNMINTPRIQTINO OPPICCS 'IOTA 4tO IOO

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Docket No. 50-335

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1975 Distribution NRC PDR Local PDR Docket File LWR 1-3 File RCDeYoung FSchroeder AKenneke RWKlecker OELD-IE (3)

HRood

'%Wilson TR Branch Chiefs LWR 1 Branch Chiefs JPanzarella ACRS (14)

Florida Power

& Light Company ATTN!

Dr. Robert E. Uhrig Vice President of Nuclear Affairs P. 0. Box 3100 Miami, Florida 33101 Gentlemen:

Before wa can complete our review of your ECCS re-evaluation for St. Lucio 1, which was docketed oa,April 24, 1975, additional information is required.

The specific information requested is contained in the Enclosure and concerns boron precipitation effects on long term cooling, the single failure criterion, aad idle loop operation.

These items have baca discussed previously with your representatives.

To maintain our review schedule, we will need your response to the items in the Eaclosure by July 3, 1975. If you canaot provide this informatioa by this date, please inform us withia seven (7) days after receipt of this letter so that wa may revise our schedule.

Sincerely, Original Signed by, Ohn Parr Olan D. Parr, Chief Light Water Reactors Prefect Branch 1-3 Divisioa of Reactor Licensiag

Enclosure:

Request for Additional Information cc:

See page 2

orrIce~

4URNAMC~

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""'""'"'"""""'""'""I'3 HRood:mck 6/ &75 L~Q3 ODParr 6/5 /75 FOIRI Al!C-3I 8 (ROT. 9.33) AECM 0240 0 U, 4. OOVeRNMCNT PRINTINO OPPICel IPN 444 l44

Florida Power

& Light Company JlJH p 3 1975 cc:

Jack R.

Newman, Esq.

Lowenstein, Newman, Reis

& Axelrad 1025 Connecticut Avenue, N. V.

Qashington,

.D. C.

20036 Norman A. Coll, Esq.

McCarthy, Steel, Hector

& Davis 14th Floor, First National Bank Building Miami, Florida 33131 Mr. John L. McQuigg P. 0. Box 1408 Stuart, Florida 33494

ENCLOSURE ST.

LUCIE l DOCKET NO. 50-335 REQUEST FOR ADDITIONALINFORHATION The ECCS provided for St. Lucie l generally satisfies our requirements with regard to long term cooling.

However, the system configurations have not been specifically evaluated to show that significant changes in chemical concentrations would not occur during the long term after a LOCA.

These potential changes in chemical concentrations have not been specifically addressed by appropriate operating procedures.

Accordingly, you should review the system capabilities and operating procedures that St. Lucie l has adopted to assure that boron precipitation would not compromise long-term core cooling capabi1ity following a loss-of-coolant accident (LOCA).

This review should consider all aspects of the St. Lucie l design, including component qualification in the LOCA environment, in addition to a detailed review of your operating procedures.

You should examine the vulnerability of your design to single failures that would result in any significant boron precipitation.

You skuld submit this,evaluation and associated operating trocmu s to us by Ze date soccer ied in the cover letter.

th -~~

moistly ~worm you about ¹ a"ceptability of any

'n c~wm~ mod> ~ications to your op~ting procedures.

These m"'-'~ic tions skuld then be promptly efze"ted to assure that boron mcipiation ~i'ot interfere w'th the ability of your fatty to con orm to Criterion (5) of 10 CFR 50.46(b).

Sol.<e concentrations may be subject to satisfactory control tom~~ opiatic~ procedures.

However, i equipz~~z modifications

~ a~ rec~~ or ¹s~le to simplify such procedu ', you should

- submit a plan which accomplishes installation of su h equipment vari.t~~

six zan~~w from the date of the cover letter.

In mmfo~~g the evaluation of single= fai1ures of ECCS eqt>>a-.en r qu~~d by Ap~diz K to 10 C:.R 50 Section'I.D.l th effects o a single failure or operator error that causes any r;an~ay-con~lied, electrically-orated valve to mave to a position that could adversely afz ct the ECCS est be

,considered.

Ou. revi w of your ECCS submittal indicates that JUN p 3 ~y5

you have rot a¹quately addressed this oonsid~mation.

Pr~~~~gly, we request dat you ciao additions"

='=ox-.a.-on addr ssh~~ the single failu~ criterio"..

~les o" valves, t;"zt ray susceptible to a single include safety injection tank isolat'on va ves,

'a~ves

"~ted 1n the pip~> frQm the r fuelir~~ wa ~~~,

-'o~a"-

+ ~'x, and tb containnent surrg to the sa="ety

""'lgect3.0.1 CURDS > GIld valves 1I1 the crossover'3Jle between the low assure and the high pressure safety injec.ion.

GL-s Reference to an idle loop operation analysis that has been based on the Commission s Interim Acceptance criteria, which are no longer applicable, is unacceptable.

Before idle loop operation with St. Lucie 1 will be permitted, you must

'provide an applicable analysis of idle loop operation based on the Acceptance Criteria for ECCS published in 10 CFR Pavt 50 on January 4, 1974.

Provide justification for the following input param t~ used in the ECCS evaluation andel:

A.

Net Free Containment Volume Discuss how the net free contaixorent volum was determined, including the consewatisms in the method.

Discuss the uncertainty associated with determining the net free contain-ment. Volum B.

Passive Heat Sinks Provide a detailed tabulation of the passive heat sinks within the contairunent (for guidance see Branch Technical Position CSB 6-1, "Minimum Containment Pressure Model for PWR ECCS Performance Evaluation", which was fovwarded to you by letter dated March 20, 1975).

Justify the heat sinks selected.

Provide a tabulation of heat sink surface ~a, materials of construction, and material thicknesses and themnphysical properties.

Discuss how heat sink surface areas were determined and the associated uncevtainty.

C.

Initiation Times of Containm nt Coolin S stems Justify that the initiation times assumed in the analysis for the containnent heat vennval systems represent the earliest possible initiation times.

JUN p g )y5

D.

Containment Initial Conditions Justify the initial values of contairunent temperature,

'pressure, and ~lative humidity used in the analysis.

Compare the initial values selected with the range of values that will be permitted duv3Jlg plant opevB.t3.one E.

Containnent S ra Water T eratuve Justify that the containment spray water temperature is the minimum temperatuze consistent with plant operating conditions.

I F. Content Fan Cooler Performance Provide the fan cooler heat reacval rate as a function of contairurent atmosphere temperature based on the minimum service water temperature consistenct with plant operating conditions.

6.32 In order for your ECCS break spectrum to be considered

complete, you must provide the results of the analyses of a break in the suction line and a hot leg break.

JUN 0

8 19yg

BRANCH TECHNICAL POSITION EICSB 18 APPLICATION OF THE SINGLE FAILURE CRITERION TO MANUALLY-CONTROLLED ELECTRICALLY-OPERATED VALVES A.

BACKGROUND Where a single failure in an electrical system can result in loss of c pability to perform a safety function, the e.feet on plant safety must be evaluated.

This is necessary, rega~d-

'C~

i ~>>>> of ":h~ h;r

"~ io"s o

finct'ion i s "aused bv a cc~ponar ma'I

> Ing to oer o..l requisite mechanical motion, or by a component performing an und sirable mechanical motion This position establishes the acceptability of disconnecting power to lectrical components of a fluid system as one means of designing against a single fai'lure that right 'cause an un-desirable component action.

These provisions are based on the assumption that th compon nt is then equivalent to a similar component that is not designed for electrical opera ion, e.g.,

a valve that can be opened or closed only by direct manual operation of the valve.

They are also based on the assumption that no single failure can both restore power to the electrical system and cause mechanical motion of the components served by the el ctrical system.

The validity of these assumptions should be verified when applying this position.

P

'I B.

BRANCH TECHNICAL POS ITI'ON 1.

Failures in both the "fail to function" sense and the "undesira'ble function" sense of components in electrical systems of valves and other fluid system compos nts should be considered in designing against a single failure, even though the valve or other fluid system component may not be called upon to function in a given safety oporational sequence.

1 2.

Where it is determined that failure of an electrical system component can cause undesired nechanical motion of a valve or other fluid system component and this motion results in loss of the system safety function, it is accept ble. in lieu of design changes that also may be acceptable, to disconnect powe~ to the electric systems of the valve or other fluid system component.

The plant technical specifications shou.d includ a list of all electrically-operated

valves, and the requir d positions of thos

valves, to

> hich the requirement for removal of electric power is applied in or"er to satisfy the single failure criterion.

Electrically-operated valves that are classified as "active" valves, i.e., are required to open or close in various safety system operatioral sequences, but are manJally-controll d, should be operated from the rain control room.

S ch valves may not be included among those valves from which power is removed in order to..eet he single failure criter on unless:

(a) electrical power can be r stored to tha valves from th-main control room,(b) valve operation is not necessary for at least ~ minutes following occurrence of '.ha event requiring such operation.

an"(c} 'it is del.onstrated 7A-?7

i" Docket Nos.60-335 w and 60-389 MAR. 20 1975 Distribution:

NRC PDR Local PDR Qg-1-3 Reading LWR 1-3 File (2)

RCDeYoung VAMoore RABirkel ACRS

('1'.4)

TR Branch Chiefs TBAbernathy, DTIE FSchroeder AKenneke-RWKlecker NWilliams ELD IE (3)

HRood VHWilson LWR 1 Branch Chiefs JRBuchanan, ORNL Florida Power and Light Company ATTN:

Dr. Robert E. Uhrig Vice President of Nuclear Affairs P. 0.

Box 3100 Miami, Florida 33101 Gentlemen:

In our letters of January 8 and 10, 1975, we identified the Emergency Core Cooling System (ECCS) evaluation as one of the outstanding items in both the St. Lucie 1 and St. Lucie 2 reviews.

To further define the scope of effort required in this'area, Enclosure 1 is attached.

The information requested therein should be included in your ECCS submittal for both St. Lucie 1 and 2. It is our understanding that the information requested in Enclosure 1 is normally part of the ECCS analysis.

Accordingly, this request should not impact your schedule for submittal of the ECCS evaluation.

Based on the currently available information. please advise us within seven (7) days of receipt of this letter, of the date by which you will submit the complete 10 CFR 50.46 analysis f'r both Units.

If you have any questions regarding this matter, please contact us.

Sincerely, Original Signed b'

0 D. Parr Olan D. Parr. Chief Light Water Reactors Project Branch 1-3 Division of'eactor Licensing

Enclosure:

Request for Additional Information cc:

See page 2

RL:LWR 1-3 OPPICC3P HRood:pga~

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MAR. 2 0 IS75 Florida Power and Light Company cc w/encl:

Jack R.

Newman, Esq.

Lowenstein,

Newman, Reis 8 Axelrad 1026 Connecticut Avenue, N. H.

Washington, D. C.

20036 Norman A. Coll, Esq.

McCarthy, Steel.

Hector and Davis 14th Floor. First National Bank Building Miami, Florida 33131 Martin H. Hodder, Esq.

1130 N. E. 86 Street Miami, Florida 33138 Mr. John L. Mcquigg P. 0.

Box 1408 Stuart, Florida 33494 OFFICC~

dVANAMCW OATC~

Form hEC.318 (ReT. 9 $3) hECM 0240 4 ll dr OOVCANMCNTFAINTINOOFFICCI IOT4 d24 Ido

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~E UEST FOR ADDITIONAL INFOR~~fATION ST.

LUCIE PLANT UNITS 1 & 2 DOCKET NOS. 50-335/389 6.27 Section 50.34 of 10 CFR Part 50 requires that an analysis and evaluation of ECCS cooling performance following postulated loss-of-coolant accidents be performed 'in accordance with the requirements of Section 50.46.

Appendix K, "ECCS Evaluation Models," to 10 CFR Part 50 sets forth cer-tain required and acceptable features of evaluation models.

Appendix K

states, in part, that the containment pressure used for evaluating cooling

,effectiveness during reflood and spray cooling shall not exce d a pressure calculated conservatively for this purpose.

It further requires that the calculation include the effects of operation of all installed pressure reducing systems and processes.'ranch Technical Position CSB 6-1,-

"Minimum Containment Pressure Model for P'de ECCS Performance Evaluation,"

V which is attached, provides additional guidance and should "be used'hen the analysis is performed.

Therefore, state the minimum containment I

t pressure that has been used in the analysis of the emergency core cooling system.

Justify this value to be conservatively low by describing the 4

conservatism in the assumptions of initial containment conditions, model-ing of the containment heat sinks, heat transfer coefficients to the heat sinks, heat sink surface

area, and any other parameter assumed in the analysi.s.

Provide the containment pressure, temperature and sump tempera-ture response for the most conservative assumptions.

MINIMUMCONTAINMENT PRESSURE MODEL FOR PWR ECCS PERFORMANCE EVALUATION A.

BACKGROUND Paragraph I.D.2 of Appendix K to 10 CFR Part 50 (Ref. 1) requires that the containment pressure used, to evaluate the. performance capability of a pressurized water reactor emergency core cooling system does not exceed a pressure calculated conservatively for that purpose.

It further requires that the calculation include the effects of operation of all installed pressure-reducing systems and processes.

Therefore, the following Branch Technical Position has been developed to provide guidance in the performance of a minimum containment pressure analysis.

B.

BKSCH TECHNICAL POSXTION 1.

Xn ut Xnformation for Model a.

Xnitial Containment Internal Conditions The minimum containment gas temperature, minimum containment

pressure, and maximum humidity that may,be encountered, under limiting normal operating conditions should be used.

b; Initial Outside Containment Ambient Conditions A reasonably low ambient temperature external to the containment should be used.

c.

Containment Volume The maximum net free containment volume should be used.

This maximum free volume should be determined from the gross contain-ment volume minus the volumes of internal structures such as walls and floors, structural steel, major equipment, and piping.

The individual volume calculations should reflect the uncertainty in the component volumes.

2.

Active Heat Sinks a.

S ra and Fan Coolin S stems The operation of all engineered safety feature containment heat removal systems operating at maximum heat removal capacity;

2 0

i.e., with all containment spray trains operating at maximum flow conditions and all emergency fan cooler units operating, should be assumed.

Xn addition, the minimum temperature of the stored water for the spray cooling system and the cooling, ~ater supplied to the fan coolers, based on technical specification limits, should be assumed.

Deviations from the foregoing will be accepted if it can be shown that the worst conditions regarding a single active failure, stored water temperature, and cooling water temperature have been selected from the standpoint of the overall FCCS model.

b.

Containment Steam Mixin With S illed ECCS Water The spillage of subcooled ECCS water into the containment, provides an additional heat sink as the subcooled ECCS water mixes with the steam in the containment.

The'ffect of "the steam-water mixing should be considered in the containment pressure calculations.

c.

Containment Steam Mixin'ithWater Prom Ice Melt The water resulting from ice melt in an ice condenser containment provides an additional heat sink as the subcooled water mixes

'ith the steam while draining from the ice condenser, into the I

lower containment volume.

The effect of the steam-water mixing should be considered in the containment pressure calculations.

2.

Passive Heat Sinks a.

Identification The passive heat sinks that should be included in the containment evaluation model should be established by identifying those structures and components within the containment that could influence the pressure response.

The kinds of structures and components'hat should be included are listed in Table 3..

l An en>

ope of passive heat sink data has been developed, based on information obtained from safety analysis reports, that would be acceptable for use in performing minimum containment pressure analyses until such time (i.e., at the OL review) that a complete identification of available heat sinks can be made.

This simplified approach has been followed on operating plants by licensees en-gaged in performing minimum containment 'pressure analyses to comply with Section 50.46 of 10 CFR Part 50.

For such cases, and for CP reviews where a detailed listing of heat sinks within the con-tainment cannot be provided, the following pxoceduxe may.be used to model the passive heat sinks within the containment:

(1)

Use the surface area and thickness of the primary containment

steel shell or steel liner and associated anchors and

concrete, as appropriate.

(2), Estimate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assume an ave'rage thickness of 3(8 inch.

(3)

Model the internal concrete structures as a slab with a

. thickness of 1 foot'nd an exposed surface area of 160,000 ft.

The heat sink thermophysical properties that would be acceptable are shown in Table 2.

At the OL stage, applicants must justify the claims they have made for heat sinks.

b.

Heat Transfer Coefficients The following conservative condensing heat transfer coefficients for heat transfer to the exposed passive heat sinks during the blowdown and post-blowdown phases of the loss-of-coolant accident should be used (See Figure 2):

(1)

During the blowdown phase, assume a linear increase in the condensing heat transfex'oefficient from h

~8 Btu/hr-ft F.

2 initial at t=0, to a peak value four times greater than the maximum calculated condensing heat transfer coefficient at the end of blowdown, using the Tagami correlation (Ref. 2),

max where hmax

0. 62 Vt P

maximum heat transfer coefficient, Btu/hr-ft -'F 2

primary coolant energy, Btu net free containment volume, ft3 time interval to end. of blowdown, sec.

( )"

During the long-term stagnation phase of the accident, (2)"

Du characterised, by low turbulence in the containment atmosphere,,

assume condensing heat transfer coefficients 1.2 times greater than that predicted by the Uchida data (Ref. 3) given in Table.

30 (3)

During the transition phase of the accident between the

'end of blowdown andi the long-term post-blowdown'phase-a p ase-,

a reasonably conservative exponential transition in the condensing heat transfer: coefficient should be assumed (See Figure 2).

The calculated condensing, heat, transfer coefficients based on the above method should be. applied ta aU. expo'sed.passive heat sinks, both metal and concrete, and for both painted.and unpainted surfaces.

F Heat transfer between adjoining materials in passive heat'inks should be b se ased'on the assumption of no resistance to heat flow at the material interfaces.

An exampl f thi h

e o s xs t e containment liner to concrete interface.

C.

REFERENCES 1.

10 CFR 50.46 "Acce t

~

j p ance Criteria For Emergency Core Cooling Systems For Light (Cater Nuclear Power Reactors" and 10 CFR P 50, art 50, Appendix K, "ECCS Evaluation=Models".

2.

T. Ta ami g

, "Interim Report On Safety Assessments and Facilities Establishment Project In Japan For Period Ending June 1965 (No. 1}"

prepared for the National Reactor Testing Station, February 28, 1966 (unpublished work).

3.

H. Uchida, A. Oyama, and Y. Toga, "Evaluation of Post-Incident Cooling Systems of Light-Mater Pmer Reactors",

Proc. Third Inter-national Conference on the Peaceful Uses of Atomic Energy, Uolume 13, Session 3.9, United Nations, Geneva (1964).

TABLE 1 IDENTIFICATION OF CONTAINMENT HEAT SINKS 1.

Containment Building (e.g., liner plate and external concrete walls, floor and sump, and liner anchors) 2.

Containment Internal Structures (e.g., internal separation walls and floors, refueling pool and fuel transfer pit walls, and shielding walls) 3.

Supports (e.g., reactor vessel, steam. generator,

pumps, tanks, ma)or components, pipe supports, and storage racks) 4.

Uninsulated Systems and Components

.(e.g.,

cold water systems, heating ventilation and air conditioning systems, pumps, motors, fan coolers, recombiners, and tanks) 5.

Miscellaneous Equipment (e.g., 1adders,.gratings, electrical cable A

trays and cranes)

TABLE 2 HEAT SINK THERMOPHYSICAL PROPERTIES Material Concrete Steel Density lb/ft~

490 Specific Heat Btu/1b-F

0. 156 0.12

. Thermal Conduc tivity

, Btu/hr-ft-F

0. 92

27. 0

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Mass Ratio

/

2.3 Heat Transfer Coefficient Btu hy ft2 F) 29 37 18 1.8 46 10 10 1 3.

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