ML17321A372

Summary of 841213 Meeting W/Util Re Cycle 6 Reload Review & Cycle 5 Kz Curve.List of Attendees & Viewgraphs Encl

ML17321A372
Person / Time
Site:	Cook
Issue date:	12/19/1984
From:	Wigginton D Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8501070604
Download: ML17321A372 (47)

Text

December 19, 1984

~ Docket No. 50-316 FACILITY:

LICENSEE:

REGARDING:

Donald C.

Cook Nuclear Plant, Unit No.

2 Indiana and Michigan Electric Company

SUMMARY

OF MEETING WITH INDIANA AND MICHIGAN ELECTRIC COMPANY REGARDING CYCLE 6 RELOAD REVIEW AND CYCLE 5 KZ CURVE

, The staff met with the licensee on December 13, 1984 to discuss the Cycle 6

reload review and schedule and to update the status on resolution of application of the K

curve to remain within the ECCS analysis throughout the remainder of cycle 5.

The licensee was supported in the discussion by Exxon Nuclear Corporation.

The list of attendees is attached as Enclosure 1.

The licensee's presentation was made from the viewgraphs in Enclosure 2.

The licensing basis discussion centered around the General Design Criteria of 1967 and 1971 as the basis for some decisions to limit the reviews or requirements on plant transients performed today.

The staff prefers to review the plant transients as we understand them today and to consider technical arguments for limiting analyses as the need arises.

The primary concern discussed was the consideration of loss of offsite power in conjunction with the limiting transient and the transient which becomes limiting because of loss of offsite power.

In this regard, the Exxon Nuclear Corporation presented a discussion of the disposition of events for the Cook 2 reload for cycle 6.

The preliminary disposition is included in Enclosure 3.

The staff's position is that those events which are BWR events do not apply and those other events listed as not-in-the-licensing basis (NILB), should be presented as not applicable or limiting on a

technical basis.

The licensee also updated the discussion of the K

curve.

Additional plots have been made of peaking factors and these will 3e presented formally in a letter from the licensee (Letter dated December 7,

1984 and subsequently received by NRRJ.<<The licensees projection still shows the current KZ curve with no derating. of plant power until mid April 1985.

The licensee's schedule for resolution was also discussed.

/s/DLWigginton

Enclosure:

As stated David L. Wigginton, Project Manager Operating Reactors Branch 81 Division of Licensing cc w/enclosure:

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MEETING

SUMMARY

0 ISTR I BUT ION N

1 Oocket or Central File NRC PDR Local PRO QRBdl Rdg J. Partlow (Emergency Preparedness only)

Steve Varga Project Manager OELD E. Jordan, J.

N. Grac'e ACRS (IO)

NSIC Gray File Plant Service List CParrish NRC Partici ants

Indiana and Michigan Electric Company Donald C.

Cook Nuclear Plant, Units 1 and 2

cc:

Mr.

M.

P. Alexich Vice President Nuclear Engineering American Electric Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43215 Attorney General Department of Attorney General 525 West Ottawa Street Lansing, Michigan 48913 Mr. Wade Schuler, Supervisor Lake Township

Baroda, Michigan 49101 W. G. Smith, Jr., Plant Manager Donald C.

Cook Nuclear Plant Post Office Box 458 Bridgman, Michigan 49106 U.S. Nuclear Regulatory Coranission Resident Inspectors Office 7700.Red Arrow Highway Stevensvi lie, Michigan 49127 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.

Washington, DC 20036 Honorable Jim Catania, Mayor City of Bridgman, Michigan 49106 U.S.

Environmental Protection Agency Region V Office ATTN:

EIS COORDINATOR 230 South Dearborn Street

Chicago, IL 60604 Special Assistant to the Governor Room 1 - State Capitol Lansing, Michigan 48909 The Honorable Tom Corcoran United States House of Representatives Washington, DC 20515 James G. Keppler Regional Administrator - Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen El lyn, IL 60137 J. Feinstein American Electric Power Service 1 Riverside Plaza

Columbus, Ohio 43216 William J.

Scanolon, Esq.

2034 Pauline Boulevard Ann Arbor, Michigan 48103 Mr.

Don van Farowe, Chief Divi s ion of Radi ol og ica l Hea 1 th P. 0.

Box 30035 Lansing, Michigan 48909 George Bruchmann Environmental and Occupational Health Services Administration Michigan Department of Health 3500 N. Logan Street P.O.

Box 30035 Lansing, Michigan 48909 Maurice S. Reizen, M.D.

Director Department of Public Health Post Office Box 30035 Lansing, Michigan 48109

Enclosure 1

D. C.'OOK FUEL RELOAD MEETING DECEMBER 13, 1984 LIST OF ATTENDEES Name D. L. Wigginton M. F.

Kennedy M. Dunenfeld R., C. Jones Jack Guttman Jim Feinstein

.Fred Adams Vance VanderBurg Bob Copeland Brian Sheron Hussein Y. Fouod George John T. A. Baxter G. F. Owsley H.

G.

Shaw R.

G. Vasey K. E.

Johnston Mike Cleveland Or anization NRR-ORB-1

ENSA, INC NRC/CPB NRC/RSB NRC/RSB AEPSC ENC AEPSC ENC NRC/RSB AEP/NSL AEPSC/Fuels

Shaw, Pittman, Potts, Trowbridge Exxon Nuclear Exxon Nuclear AEPSC NRR-ORB-1 AEPSC

Enclosure 2

AGENDA

~

~

ENC/AEP/NRC HEETING ON DECPlBER 15.

1984 I'n RoouI:Vro LICV<Sr.<Ii SOSIS REYIEM OF DRAFT DISPOSIT'ION OF EYENTS 1

STATUS AND SCHEDULE FOR ~RLANT Ti~RANSIENT ANALYSES

~

~

PLANT STATUS WITH RESPECT TO FhH LINIT ENC LOCA AXIAL SHAPES RESULTS ENC ENC AEP ENC SUN lARY AEP

DOCUHENTS EVALUATED

'(Applicable sections of the folio@i d

e o

ov ng documents mre reviewed)

Vpdat~ ZSAR 2.

Original FShm Questions 3.

Xaitial SKH, September 10,

$ 974 4.

sz SEB Suppleaent 7, December 23, 1977 5.

Federal Register, July

$ 1,

$ 967 February 20, t97 5 July 7>

$ 971 February 12,

$ 976 October 27,

)978

>.4 ERAL DESIGN CRITERIA The general design cxiteria followed in the design of this plant have been developed as performance criteria which define or describe safety objectives and pxocedures, and they provide a guide to the type of plant design inforiation which is included in this report.

These criteria are specifically addressed in the chapters of the PSAR where they are pertinent.

An index to the criteria is given in Table 1 ~ 4-1.

In the chapter where a specific criterion is relevent to the design, the cziterion is quoted and is followed by a brief summary of the design or procedures.

The design or procedures are then more fully described in other sections of the chapter.

Othoz criteria which 1

app y generally to the design of the plant are given in Section 1 ~ 4.1 ~

In addition, the Donald C. Cook Nuclear Plant has been designed to comply with the AppU.cant's understanding of the intent of the AEC proposed General Design Criteria, ai published for comment by the AEC in July, 1967.

The application of the AEC proposed General Design (1)

Criteria to the Donald C. Cook Nuclear Plant was discussed in the original FSAR, Appendix H.

Table 1 ~ 4-1 cdntains a cxoss-index between the AEC design criteria and the FSAR chapters where those criteria are intezpreted.

1 ~4 1

ali Standards Criterion:

Those systems and components of reactor facilities which are essential to the prevention, or the mitigation of the consequences, of nuclear accidents which could cause undue risk to the health and safety of the public shaU. be identi-fied and then designed, fabricated, and erected to quality standards that reflect the importance of e safety function to be perforeed

@here generally recognized codes and standards pertaining to design, zhaterials, fabrication, and inspection are used, they shall be identified.

Mhe'xe adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessazy" 1 ~ 4-1 July>

1982

CA 3.0 DESIGH CRITERIA STRUCTURES COMPONENTS UIPHKHT AND SYSTEMS 3.1 Conformance with AZC General Desi Criteria The Cook plant vas designed and constructed to meet the, intent of the Proposed General Design Criteria, published July ll, 1967.

The Final Safety Analysis Report had been filed with the Commission vhen revisions of the General Design Criteria vere published in February 1971'nd July.7, 1971.

We reviewed the plant design against the current General Design'Criteria and ve believe that the design meets these criteria.

3.2 Classification of Structures onents and S stems The applicant has classified the plant structures, components, and systems into three principal categories.

Category* I includes those structures, components, and systems whose failure might cause or increase the severity of a loss-of-coolant accident, or result in an uncontrolled release of aigaificant amounts of radioactivity and those structures, components, aad systems essential to safe reactor shutdown.

Category II includes those structures, components, and systems that are iaportant to reactor operatioa, but not essential to safe reactor otutdova aad vbooe failure weald aot result in the release"of oigaificeat enounts of radioactivity.

Category IIIincludes those structures,

systems, and ccepcmeats that are not directly related to reactor operation or containment.

• In the FSAR, "Category" is termed "CLass."

36 FEDERAL REGISTER 3255 AND 3256 GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS The Atomic Energy Commission has adopted an amendment to its regulations 10 CFR Part 50, "Licensing of Production and Utilization Facilities," which adds an Appendix A, "General Design Criteria for Nuclear Power Plants."

The "General Design Criteria for Nuclear Power Plants" Mded as Appendix A to Part 50 establish the minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission.

They also provide guidance In establishing the principal design criteria for other type~

of nuclear power plants.

Principal design criteria established by an applicant and accepted.by the Commission will be incorporated by reference in the construction permit.

In considering the issuance of an operating license under Part 50, the Commission will require assurance that these criteria have been satisfied in the detailed designed and construction of the facility and that any changes in such',

criteria are Justified.

APPLZCSBELZTX'F QDC We QDC de@ca~

5m the PSRB ~ilp te tt/he Ronald C. Ccrc Ehaelbeaa'3aett, Sutitt; 2.

Page

'4 EVENTS TO BE CONSIDERED Those events which vere analyzed at the time the*..

Donald C.Cook Nuclear Plant, Unit 2.Operating License was issued, constitute the events in the licensing basis for-this plant.

Page 5

DXSPOSITlON OF EVENTS Those events -which.are described in the original FSAB Mll be evaluated in performing the disposition of events described in the

<Exxon Nuclear Methodology for Pressurized Hater Reactors:

Analysis of Chapter 15 Events.~

This disposition sill be documented and vill state which <meets are to be r eanalyzed, which are bounded, which Mll reference previous analyses, and which Standard Reviev Plan Events. are not in cer 1icem~iag basis..

Page 6

EVENTS TO BE EVALUATED 14.1.1 Uncontrolled Rod Cluster Control Assembly Bank Mithdrawal from a subcritical oondition.

14.1.2 Uncontrolled Rod Cluster Control Assembly Bank Mithdraval at Poser.

10.1.3 Rod Cluster Control Assembly Misalignment 14.1.4 Rod Cluster Control Assembly Drop 14 ~ 1.5 14.1.6 Uncontrolled Boron Dilution Loss of Forced Reactor Coolant Floe 14.1.6.1 Single Reactor Coolant Pump Looked Rotor

14. 1.7 Startup of an Inactive Reactor Coolant Loop

14. 1. 8 14 ~ 1.9 Loss of External Electr'ical Load and/or Turbine Trip Loss of Normal Feedmter 14 1.10 Excessive Heat Reeaval Due to Feedvater System Malfunctions 14.1.11 Excessive Load Xncrease Xncident 14 1.12 Loss of Off Site Povor to the Station Auxiliaries (Station Blackout)

14. 1. 13 14.2.1 Turbine - 6enerator Accident Fuel Handlins Accident 14.2,2 Accidental Deloaso of Radioactive Liquids 14.2.3 1%.2.O 14.2.5 14.2.6 Accidental Haste Oas Release Steam Qamrntor Tube Ruptur e Sup~ of n Stcaa Line Rupture of a Control Rod Drive Mechanism Housing 14.2.8 Ha~ Rupture of a Hain Feedmter Pipe

Page 7

Ma+r LOCA LOCA froa a Saall Ruptured Pipe or f'roe Cracks fn a Large Pipe which Actuatea the ECCS Fuel Cask Drop

Page 8

LOSS OF OFF SITE POMER E

t h

h require reanalysis using the Exxon Nuclear Company ven svac methodology vill be explicitly analyzed vith and vxthout ooff site pover if the event vas so analyzed in the FSAR.

Page 9

EXPLICIT LOSS OF OFF SITE FOYER BVENTS 14.1.9 Loss of Nor mal Feedmter 14.1 '2 Loss of Off Site Poser to Station kuxiiiaries (Station Blackout) 14.2.4 Steam Generator Tube Rupture 14.2.5 Rupture of a Steam Line 14.2.6 Rupture of a Control Rod Drive Hec4udsa Housing (RCCi E+ction) 14.2.8 HaJor Rupture of a Hain Feedmter Pipe 14.3. 1

~or LOCi 14.3.2 Loss of Reactor Coolant froa a Saall Ruptured Pipe or free a Crack in a Large Pipe vhich ictuates the ECCS.

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PLANT TRANSIENT EVENT DISPOSITION WILL BE DONE IN ACCORDANCE TO ENC GENERIC PROCEDURE IN XN-NF-84-73(P)

HILL USE D, C.

COOK UNIT 2 PLANT CONFIGURATION HILL USE 196?

FEDERAL REGISTER GDCs AS GIVEN IN THE D, CD COOK UNIT 2 FSAR AS LICENSING BASIS

PRELIMINARY DISPOSITION OF EVENTS

~

IDENTIFIES EVENTS FOR REANALYSIS FOR CYCLE 6,

~

REVIEW TODAY TO ELICIT STAFF RESPONSE TO PROPOSED ANALYSIS SCOPE.

P PLANT TRANSIENT ANALYSIS SCHEDULING SCHEDULE NRC MEETING TO DISCUSS D.

C.

COOK 2 LICENSING PLANS OCTOBER 15.

1984 NRC MEETING TO PRESENT PRELIMINARY DISPOSITION OF EVENTS SUBMIT FINAL DISPOSITION OF EVENTS'RC f",EETING TO REVIEW FINAL DISPOSITION DECEMBER 13.

1984 MARCH 1985 MARCH 1985 SUBMIT DRAFT APPENDIX C

(PRESCRIPTION OF EVENTS FOR 4-LOOP PLANTS)

MAY, 1985 NRC MEETING TO REVIEW. STATUS OF ANALYSES MAY 1985 PARTIAL SUBMITTAL OF D.

C.

COOK UNIT 2 CYCLE'6 ANALYSES SUBMITTAL OF REMAINING D.

C.

COOK UNIT 2 ANALYSES PLANT SHUTDOWN PLANT STARTUP JULY 1985 AUGUST 1985 OCTOBER 1985 DECEMBER 1985 4

MOS.

AXIAL SHAPES FOR LOCA ANALYSIS IN NOVEI'1BER 13.

1984 MEETING.

ENC PROPOSED ESTABLISHING THE SHAPES BASED ON AXIAL OFFSET REQUIRENENTS ON F~H FOR THIS NEETING.

AEP WILL GIVE PLANT STATUS AND ENC WILL GIVE THE ANALYSIS STATUS

INTERIf'1 NETHODOLOGY TECHNICAL SPECIFICATION L INITS LARGE AND SHALL BREAK LOCA 1.0 K(Z)

BOTTON AXIAL HEIGHT TOP LARGE BREAK AXIAL SENSITIVITY zH L

AO

CYCLE SPECIFIC ANALYSIS DEVELOP SPECTRUN OF AXIAL POWER SHAPES VERSUS AO AXIAL SHAPES DEVELOPED AT HFP WITH XTG ANALYSIS ACCOUNTS FOR CYCLE EXPOSURE, BURNABLE ABSORBER DEPLETION, AND BORON CONCENTRATION ANALYSIS PERFORNED FOR A RANGE OF AXIAL OFFSETS EXPECTED TO BOUND CYCLE REACTOR OPERATION SHAPES REPRESENTATIVE OF OPERATION WITHIN TECHNICAL SPECIFICATION LINITS ON AXIAL OFFSET WITH RESPECT TO LOCA.

THE WORST AXIAL SHAPES ARE IDENTIFIED VERSUS AO BY SCANNING SHAPES TO DETERf'lINE THOSE WHICH HAVE THE HIGHEST LHGRs IN TOP OF CORE DETERNINE AN F~H WHICH SATISFIES LOCA LINIT

ENC ANALYSIS STATUS HAVE SELECTED AXIAL SHAPES AND BEGAN ANALYSES OF THOSE SHAPES ANALYSIS IS AN ITERATIVE.CALCULATION TO DETERMINE THE MAXIMUM F~H

1.50

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0.25

+h.Q X A. 0.

+9. 0 X A. 0.

-1.0 X A.O.

0.00 2

4 5

6 7

8 Distance from Oottom of Core (feet) 10 11 Figure 1

D.C.

Cook Unit 2 Cycle 5

DOC Hot Channel Axial Power Profile

1.50 1.25 1.00 5-O cu 0.75 0.50 0, 7")

"2.3 X A.D.

+2.7 X A.D.

"'1 0 K

A 0 0.00 0

4 5

6 7

8 Oistance from Bottom of Core (feet) 10 Figure 2

O.C.

Cook Unit 2 Cycle 5

NOC Hot Channel Axial Power Profile

1.25 1.00 O

Z 0.75 0.50 "1.8 X A.O.

0.25 "6.8 X A.O.

0.00 0

4 5

6 7

8 Distance from Hottom of Core (feet) 10 Figure 3

D.C.

Cook Unit 2 Cycle 5

EOC Hot Channel Axial Power Profile

Enclosure 3

PRELIMINARY DISPOSITION OF EVENTS FOR D'.C.

COOK UNIT 2 CYCLE 6 Category I:

Increase in Heat Removal b

Secondar S stem SRP Event Number 15.1.1 15.f.2 15.1.3 15.1.4 15.1.5 Event Decrease in Feedwater Temp.

Increase in Feedwater Flow Hot Full Power H7P Case Increase in Steam Flow, Inadvertent Opening of. an S.G.

Relief or Safety Valve Steam System Piping Failures Bounded Bounded (15.1.3)

Bounded (15.1.3)

Bounded (15.4.1)

Analyze Bounded (Current Analysis Bounded (Current Analysis)

Category II:

Decrease in Heat Removal b

Secondar S stem 15.2.1 15.2.2 15.2.3 15.2.4 15.2.5 15.2.6 15.2.7.

15.2.8 Loss of External Load TUrbine Trip Loss of Condenser Vacuum Closure e of MSI V Analyze Bounded (15.2.1)

Bounded (15.2.1)

NILB* (BMR)

Loss of Non-Emergency A.C. Power Loss of Normal Feedwater Feedwater System Pipe Breaks Bounded (15.3.1, 15.2.?)

Analyze Analyze as loss of 1 S.G. with SLOTRAX Steam Pressure Regulatory Failure NILB

• NILB = Not in Licensing Basis.

SRP Event Number Event Bounded Category III:

Decrease in Reactor Coolant S stem Flow Rate 15.3.1 15.3.2 15.3.3 15.3.4 Loss of Forced Reactor Coolant Flow Flow Controller Malfunction Reactor Coolant Pump Rotor Seizure Reactor Coolant Pump Shaft Break Bounded (15.3.3)

NILB* (BWR)

Analyze (For DNB only)

Analyze (For DNB only)

Category IV:

Reactivit and Power Distribution Anomalies 15.4.1 15.4.2 15.4.3 15.4.4 15.4.5 15.4.6 15.4.7 15.4.8 Uncontrolled CRA Withdrawal from Subcritical Uncontrolled CRA Withdrawal Power Control Rod Misoperation Rod and Bank Drop Static Misalignment Single CRA Withdrawal Startup of an Inactive Loop Flow Controller Malfunction CVCS Malfunction Resulting in Reduced RCS Boron Concentration Inadventent Loading of a Fuel Assembly in an Improper Position Rod Ejection Accidents Analyze Analyze Analyze Analyze NILB Bounded (Current Analysis)

NILB* (BWR)

Bounded (15.4.2)

(Not credible due to administrative controls)

Analyze

• NILB = Not in Licensing Basis.

SRP Event Number Event Bounded Category V:

Increase in Reactor Coolant Inventor 15.5.1 Inadvertent Operation of ECCS That.Increases RCS Inventory NILB*

15.5.2 CVCS Malfunction That Increases NILB RCS Inventory Category VI:

Decrease in Reactor Coolant Inventor 15.6.1 15.6.2 Inadvertent Opening of a Pressurizer PORV Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment NILB Bounded (Current Analysis) 15.6.3 15.6.4 15.6.5 Radiological Consequences of Steam Generator Tube Failure Radiological Consequences of Hain Steam Line Failure Outside Containment Loss of Coolant Accident Bounded (Current Analysis)

NILB (BWR Event; Pad Consequence of 15.15 considered there)

Analyze Category VII:

Radioactive Release from a Subs stem or Com onent 15.7.1

~

15.7.2 15.7.3 Waste.

Gas System Failure Radioactive Liquid Waste System Leak or Failure Radioactive Liquid Releases due to Liquid Containing Tank Failure Bounded (Current Analysis)

Bounded (Current Analysis)

Bounded (Current Analysis) 15.7.4 15.7.5 Radiological Consequences of Fuel Handling Accident Spent Fuel Cask Drop Accident Bounded (Current Analysis)

Bounded (Current Analysis)

• NILB = Not in Plant Licensing Basis

15.1.1 Decrease in Feedwater Heatin A decrease in FW temperature can result from bypass of k'2, 3, and 4

FW heaters upon inadvertent opening of the bypass line valve.

Complete bypass of these heaters results in a reduction of FW enthalpy by <78 BTU/lb.

The resulting increased load demand on the primary is about 2.5X less than that resulting from the lOX load increase considered in 15. 1.3.

Because event 15.1.1 and 15.1.3 are similar in respect to their impact on primary

system, and because the magnitude of the event initiator for 15.1.1 is bounded by that of 15. 1.3, this event is bounded by the increase in steam flow (15.1.3).

Rate of evolution of this event is also bounded due to S.G. thermal inertia.

15.1.2 Increase in Feedwater Flow The event is initiated by failure to full open position of a single feedwater control valve.

The increase in feedwater flow places an increased load demand on the primary which is anticipated to be bounded in magnitude and rate of occurrence by that which occurs in 15.1.3 (for HFP cases).

The event can occur at zero power also and results in an uncon-trolled reactivity insertion at EOC.

A bounding value for the magnitude of the reactivity insertion may be estimated, neglecting the thermal, inertia of the S.G. fluid.

This value is smaller than that which is used for the withdrawn control rod in the URN event from a subcritical condition (15.4.1).

The results of the increase in feedwater flow are therefore expected to be bounded by those of the URW from subcritical.

15.1.4 Inadvertent 0 enin of an S.G. Relief or Safet Valve This event can occur at full power or no load conditions.

The capacity of a single S.G.

PORV is 2.5X of total full load steam flow; the capacity of a single S.G. safety valve is 5.8X of full load. steam flow.

The effects of either valve opening inadvertently are therefore bounded by the proposed 10K load increase to be employed in the simulation of 15.1.3.

The case from no load conditions is bounded by the UFSAR analysis.

The opening of the secondary side valve imposes a load demand on the system, resulting in a cooldown of the moderator.

'At EOC conditions, with a large negative

HTC, the event results in an approach to criticality.

The controlling parameters of the event are thus the magnitude of the moderator cooldown and the magnitude of the EOC HTC.

The magnitude of the cooldown is controlled by the steam flow capacity of the opened secondary valve, which has not changed between Cycle 3 and 6.

Thus, the magnitude of the cooldown is not expected to.have changed.

The magnitude of the EOC HTC is limited by the Technical Specifications to a value more positive than -39 pcm/oF.

This technical specification will remain in effect for Cycle 6; thus, no adverse change in core kinetics response to a cooldown relative to conditions on which the reference analysis was based is expected.

The controlling parameters of the event are therefore unchanged with respect-to conditions supported by the reference analysis.

15.1.5 Steam S stem Pi in Failures The results of the main steamline break are bounded by the Cycle 4 analysis.

One changed condition exists for Cycle 6 relative to the

conditions supported in Cycle 4:

steam generator tube plugging.

The effect of tube plugging is to reduce system coolant flow and to reduce steam generator heat transfer surface area.

The reduction in heat transfer surface area will mitigate the primary system cooldown slightly and is, therefore, bounded by the unplugged case considered in Cycle 4 with respect to cooldown.

The kinetics feedback employed in Cycle 4 analysis bound both Cycle 4

and Cycle 6 cores.

Thus, the return to power calculated for Cycle 4 will be at worst unchanged for Cycle 6.

The effective flow decrease between Cycle 4

and Cycle 6.is 7X.

Sufficient conservatism was applied to the flow used in the Cycle 4 initial heat flux calculation to more bound this 7X flow decrease.

Thus, calculated MCHFR is not expected to change as a result of the flow decrease.

No fuel was calculated to experience boiling transition in the Cycle 4

MSLB; the same result would characterize the Cycle 6 calculation.

Thus, the Cycle 6 main steam line break event is bounded by the results reported for Cycle 4.

15.2.2 and 15.2.3 Turbine Tri and Loss of Condenser Vacuum The results of either event are bounded by the loss of external load event as analyzed (15.2.1).

The loss of external load event as simulated is initiated by closure of the turbine stop valve, without direct reactor trip, without available condenser

bypass, and without available steam dump system.

The turbine trip event would be initiated by closure of the turbine stop valve, but would result in a direct reactor trip.

The loss of condenser vacuum would result in turbine trip, followed by direct reactor trip on turbine trip, without available condenser bypass.

The loss of external load as simulated thus

is initiated by the fastest isolation of the turbine possible (closure of turbine stop valve),

nelects the direct reactor trip which is available for turbine trip and loss of condenser

vacuum, results in the maximum load rejection by the secondary due to assumed unavailability of the condenser

bypass, steam dump

system, and S.G.

PORVs.

15.2.6.

Loss of Non-Emer enc A.C.

Power The short term consequence of LOOP is the loss of power to the RCS pumps.

In the short term, the event is therefore essentially the same in its impact on the primary system as the four pump coastdown event.

The four pump coastdown is treated in 15.3.1.

The complete loss of offsite power is assumed for the loss of normal feedwater event.

Thus, the longer term consequences of LOOP (RCS swell and overpressurization) are treated in that event (15.2.7).

15.2.8 Feedwater S stem Pi e Break The event results from a rupture of a main feedwater pipe.

If the rupture occurs upstream of the feedline check valve, the event proceeds as a

loss of normal feedwater.

If the rupture occurs between the S.G.

and the check valve, the affected steam generator will blowdown, resulting in either primary system cooldown or primary system heatup.

The cooldown branch of the event is bounded by the main steam line break event (15.1.5).

Possible system heatup consequences will be evaluated using the SLOTRAX code.

15.3.1 Loss of Forced Reactor Coolant Flow This event is bounded with respect to calculated MDNBR by the locked rotor event-(15.3.3).

The locked rotor event is characterized by a more rapid core flow degradation, and results in lower core flows through the time of DNB for that reason.

ENC calculations have supported this conclusion for all W

plants analyzed.

If, however, the locked rotor event results in penetration of the DNB SAFDL, the four pump coastdown will be analyzed to demonstrate that condition II event acceptance criteria are met.

ENC calculations have consistently demonstrated that peak RCS pressures reached as a result of the loss of flow events are substantially bounded by. the peak pressures calculated in the loss of external load event

.(15.2.1).

The capability of the pressurizer safety valves to avert penetration of the vessel pressurization limit is, therefore, considered in event 15.2.1.

15.3.3 and 15.3.4 Pum Rotor Seizure/Pum Shaft Break Both events will be analyzed for MDNBR.

Peak pressurization cases will not be performed (see'aragraph 2, 15.3.1).

15.4.4 Startu of an Inactive Loo The Cook-2 plant is precluded by its license from operating with less than four loops in operation above the P-7 setpoint.

The event was analyzed in the UFSAR from an initial power level of 71K.

The P-7 interlock setpoint is ill of rated power.

The controlling parameters for this event are the EOC MTC and the magnitude of the core moderator cooldown occasioned by starting the inactive loop.

The EOC MTC is limited by a Technical Specification unchanged from the reference analysis.

The magnitude of the cooldown is governed by the temperature difference between the core water and the water in the inactive loop at event initiation.

The reference analysis supported a temperature difference of about 50oF; the Cycle 6 case would have to consider a temperature difference of less than 15oF.

The basis for these temperature differences is outline below.

The event for Cycle 6 is thus bounded by the reference analysis.

Reference Analysis (71K ower Cycle 6 Considering License Conditions ill ower Steam Temperature at Initial Condition

-520oF

- 539oF Inactive Loop Temperature at Initial Condition

-520oF

- 539oF Core Average Temperature at Initial Condition 566oF 550oF T driving force for Cooldown (Tcore TLOOP~

46oF lloF 15.6.2 Radiolo ical Conse uences of Failure of Small Lines Carr in Primar Coolant Outside Containment An analysis for D.C.

Cook II was submitted in Cycle 5 to assess the impact of a slight configuration change in the ECCS.

That analysis considered a peak LHGR of 16.67 kw/ft, having been structured to conser-vatively bound both Units 1 and 2.

Peak expected LHGR in the Unit 2 core is about 11.5 kw/ft, reduced from the Unit 1 value by nearly 46K by the increased number of fuel rods in the Cook II core (17x17 vs. 15xl5) and the reduced technical specification Fq limit in Unit 2.

Thus, because PCT calculated for

the small break is strongly dependent on LHGR, the analysis submitted for Cycle 5 is very conservative for the Cook II plant.

PCT for the Cycle 5

analysis was below 1750oF; quite substantial margin to limits therefore exist for the Cook II plant in a postulated small break LOCA.

The relevant physical effects of tube plugging are a roughly 3.5X decrease in total system coolant value initiallyresident above the core and a

10X decrease in steam generator heat transfer area.

These effects are minimal in magnitude, and more than offset by the 45K conservatism in peak LHGR assumed in the Cycle 5 submittal.

15.6.3 Radiolo ical Conse uences of Steam Generator Tube Failure Radiological consequences of this event have been treated in XN-NF-82-90(P),

Supp.

1.

The event results are independent of tube plugging

levels, and insensitive to fuel type.

Because technical specification limits on primary and secondary coolant activity are unchanged and no fuel failure is expected for this

event, the results of the FSAR analysis remain applicable for Cycle 6..

15.7.1, 15.72, 15.7.3, 15.7.5:

Waste Gas S stem Failure, Radioactive Li uid Waste S stem Leak or Failure, Radioactive Li uid Releases Due to Tank

Failure, S ent Fuel Cask Dro Accident These events are treated in the FSAR, responses to FSAR questions, or UFSAR for Cook Unit II.

The results are not dependent on either fuel type or on steam generator tube plugging.

The original analyses are thus not affected by the current licensing action and remain applicable.

15.7.4 Radiolo ical Conse uences of Fuel Handlin Accident The event is treated in XN-NF-82-90 and XN-NF-82-90, Supplement 1

in a manner to bound Cycle 6.

The results are not dependent on steam generator tube plugging.

D.C.

Cook Unit 2 Status of Current Anal sis

, SRP Event Number Event Bounded Category I:

Increase in Heat Removal b

Secondar S stem 15.1.1 15.1.2 Increase in Feedwater Flow UFSAR 14.1.10, and App.

14B Decrease in Feedwater Temperature UFSAR 14.1.10 15.1.3 15.1.4 15.1.5 Increase in Steam Flow Inadvertent Opening of an S.G.

Relief or Safety Valve Steam System Piping Failures UFSAR 14.1.11 UFSAR 14.2.5 UFSAR 14.2.5, 14.3.4, 14.3.7, XN-NF-82-90 Supp.

1,'OG

Studies, XN-NF-82-32(P), Rev.

1 Category II:

Decrease in Heat Removal b

the Secondar S stem 15.2.1 15.2.2 15.2.3 15.2.4 15.2.5 15.2.6 15.2.7 15.2.8 Loss of External Load Turbine Trip Loss of Condenser Vacuum Closure of MSIV Steam Pressure Regulator Failure Loss of Non-Emergency AC Power to Station Auxiliaries Loss of Normal Feedwater Feedwater System Pipe Breaks UFSAR App. 14.B UFSAR App. 14.B UFSAR App. 14.8 BWR Event UFSAR 14.1.12, 14.1.6, 14.1.9 UFSAR 14.1.9 UFSAR 14.2.8; 14.3.7

• Not in Cook II Licensing Basis.

SRP Event Number Event Bounded Category III:

Decrease in Reactor Coolant S stem Flow Rate 15.3.1 15.3.2 15.3.3 Loss of Forced Reactor Coolant Flow UFSAR APP.

14.B Flow Controller Halfunction BWR Event Reactor Coolant Pump Rotor Seizure UFSAR App. 14.B Category IV:

Reactivit and Power Distribution Anomalies 15.4.1 Uncontrolled CRA Withdrawal from UFSAR App. 14.8 Subcritical or Low Power Condition 15.4.2 15.4.3 15.4.4 15.4.5 15.4.6 15.4.7 Uncontrolled CRA Withdrawal (Power)

Control Rod Hisoperation Startup of an Inactive Loop Flow Controller Halfunction CVCS Halfunction Resulting in Reduced RCS Boron Concentration Inadvertent Loading of a Fuel Assembly in an Improper Position UFSAR App. 14.B UFSAR 14..1.3,.14.1.4 UFSAR 14.1.7 BWR Event UFSAR App. 14.8 15.4.8 Spectrum of Rod Ejection Accidents XN-NF-83-85 and Supp.

1 Rev.

1 Category V:

Increase in Reactor Coolant Inventor 15.5.1 15.5.2 Inadvertent Operation of ECCS That Increases RCS Inventory CVCS Halfunction That Increases RCS Inventory

• Not in Cook II Licensing Basis.

SRP Event Number Event Bounded Category YI:

Decrease in Reactor Coolant Inventor 15.6.1 15.6.2 15.6.3 15.6.4 15.6.5 Inadventent Opening of Pres-surizer PORV Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Contain-ment Radiological Consequences of Steam Generator Tub'e Failure Radiological Consequences of Main Steam Line Failure Outside Containment Loss of Coolant Accident UFSAR 14.3.7, 14.3.8 Cycle 5 Submittal UFSAR 14.2.4, XN-NF-82-90, Supp.

1 BWR Event XN-NF-84-21(P) Rev.

2 UFSAR 14.3.4, 14.3.6 XN-NF-82-'90, Supp.

1 Category VII:

Radiactive Release From a Subs stem or Com onent 15.7.1 15.7.2 15.7.3 15.7.4 15.7.5 Waste Gas System Failure Radioactive Liquid Waste System Leak or Failure Radioactive Releases Due to Liquid Containing Tank Failure Radiological Consequences of Fuel Handling Accident Spent Fuel Cask Drop Accident UFSAR 14.2.3, 14.3.5 UFSAR 14.2.2 UFSAR 14.2.2 XN-NF-82-90, Supp.

1 FSAR App. 70, 73 (Response to Q14.15)

Not in Cook II Licensing Basis.

~ %4 1 41V 0

0 l> I 'l %A~

~AIR 44

~ w v

~ Feah 0'

Al =

I '4 o'

4

~

SRP Number Event SRP Event Number 14.3.3 14.3.4 14.3.6 14.3.7 Core Internals Integrity Containment, Integrity Hp in Containment After LOCA Long Term Cooling 15.1.5, 15.6.5 15.6.5 15.1.5, 15.2.8, 15.6.2 14.3.8 Nitrogen Blanketing 15.6.2 Note:

UFSAR App. 14.B is Positive Moderator Temperature Coefficient Analysis.

Rev.

1

0