ML17334A814

Application for Amend to License DPR-58,revising Tech Specs to Extend Fuel Peak Pellet Burnup & Increase Fq Value Limit in Fuel.Fee Paid

ML17334A814
Person / Time
Site:	Cook
Issue date:	08/23/1984
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17334A815	List: ML17334A814 ML17334A816
References
AEP:NRC:0745M, AEP:NRC:745M, NUDOCS 8409050167
Download: ML17334A814 (130)

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1 ~

REGULATOR'r INFORMATION DISTRIBUTION SYSTEM (RIDS)

'ACCESSION NB/ 8009050167, DOC ~ DATE e 84/08/23 NOTARIZED!"'O DOCKET ¹ FACIL:50 315 Donald C ~

Cook Nuclear Power Pl anti Unf t 1, Indiana-8 05000315 AUTH~ NAME AUTHOR AFFILIATION ALEXICHiM~ P.

Indiana L Michigan Electric Co.

RECIP, NAME RECIPIENT AFFILIATION 05000315 SUBJECT! Application for amend to License OPR 58 revising Tech Specs to extend fuel peak pellet burnup L increase FQ value limit in fuels Fee paido DISTRIBUTION CODE:

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INDIANA& MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 August 23, 1984 AEP: NRC: 0745M Donald C.

Cook Nuclear Plant Unit No.

1 Docket No. 50-315 License No. DPR-58 TECHNICAL SPECIFICATION CHANGE REQUESTS Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.

C.

20555

Dear Mr. Denton:

By this letter and its attachments, we request changes to the Technical Specifications for the Donald C.

Cook Nuclear Plant Unit No.

1.

The proposed revised Technical Specification pages are contained in Attachment A.

The reasons for the proposed changes to the Technical Specif'ications, and the Justifications that the changes do not involve significant hazards considerations, are contained in Attachments B and C to this letter.

The changes described in Attachment B involve extending the peak pellet burnup in fuel supplied by Exxon Nuclear Company from 42,200 MWD/MTU (42.2 %0)/KG) to 48,000 MWD/MTU (48.0 MWD/KG).

These changes are supported by a LOCA Analysis and additional information regarding mechanical

design, which was sent to you directly by Exxon Nuclear Company with letter JCC:113:84, dated August 21, 1984.

The current burnup limit is expected to be reached on November 30, 1984.

Without this burnup extension, we would be unable to continue operation of Cycle 8

because of the requirements of Technical Specification Section 3.2.2.

The changes proposed in Attachment C and supported by Attachment D involve an increase in the F limit in fuel supplied by Westinghouse f'rom 1.97 to 2.10. It should be noted that part of this analysis is based on use of the BART code, which has not been previously used for the Donald C.

Cook Nuclear Plant, Unit 1 ~

These proposed changes have been reviewed by the Plant Nuolear Safety Review Committee (PNSRC) and will be reviewed by the Nuclear Safety and Design Review Committee (NSDRC) at their next regularly scheduled meeting.

In compliance with the requirements of 10 CFR 50.91(b)(1),

a copy of'his letter and its attachments have been transmitted to Mr. R.

C. Callen of the Michigan Public Service Commission.

8409050167 840823 PDR ADOCK 050003l5 F'

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Mr. Harold R. Denton AEP: NRC: 0745M Pursuant to 10 CFR 170.12, we have enclosed a check in the amount of $150.00 as payment for the application fee for the proposed amount.

This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, M

. Ale ich Vice Preeidect

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Mr. Harold R. Denton

~\\3 AEP: NRC:0745M Attachments:

A.

Proposed Revised Technical Specifications Pages for D.C.

Cook Unit 1.

B.

Reasons for the extension of the peak pellet burnup allowed in fuel supplied by Exxon Nuclear Company and Justification that the changes do not involve significant hazards considerations.

C.

Reasons for the increase in F~ for fuel supplied by Westinghouse.

D.

"D.C. Cook Unit 1 3411 MWt Large Break LOCA Analysis", Westinghouse Electric Corporation,

June, 1984.

cc: John E. Dolan W. G. Smith, Jr. - Bridgman R.

C. Callen G. Charnoff E.

R. Swanson, NRC Resident Inspector - Bridgman

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Mr. Harold R. Denton AEP: NRC: 0745M Attachment D

"D.C. Cook Unit 1 3411 MWt Large Break LOCA Analysis",

Westinghouse Electric Corporation,

June, 1983.>>

'This document has been piepared by Westinghouse Electric Corporation in a format foi eventual inclusion in the Donald C; Cook Nuolear Plant FSAR.

Although this is ~o intended for that purpose at this time, the format has been retained for convenience.

14.3.1.1 Major LOCA Analyses Applicable to Westinghouse Fuel Identification of Causes and Fre uenc Classification

-A loss-of-coolant accident (LOCA) is the result of a pipe rupture of the RCS pressure boundary.

For the analyses reported

here, a major pipe break

( large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 ft This event is considered an 2

ANS Condition IV event, a,limiting fault, in that it is not expected to occur during the lifetime of D.

C.

Cook Unit 1, but is postulated as a

conservative design basis.

The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 (30 CFR 50.46 and Aopendix K of 10 CFR 50 1974) as follows:

1.

The calculated peak fuel element clad temperature is below the requirement of 2,200'F.

2, The amount of fuel element cladding that reac;s chemically with water or steam does not exceed 1 percent of:he total amount of 2ircaloy in the reac or.

3.

The clad temoerature transient is terminated at a time when the core aeometry is still amenable to cooling.

The locali-ed cladding oxidation limit of 17 percent is not exceeded d ring or af er quenching.

4.

The core remains amenable to cooling during and after :he break.

5.

The core temperature is reduced and decay heat is removed for an

'I

..'xtended period of time, as required by he long-lived "radioactivity".

remaining in the core.

These criteria were established to provide significant margin in emergency core cooling system (ECCS) performance following a LOCA.

WASH-1400 (USNRC 1975) presents a recent study in regards to the probability of occurrence of RCS pipe ruptures.

Se uence of Events and S stems 0 erations Should a major break occur, depressurizaton of the RCS results in a pressure decrease in the pressurizer.

The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached'.

A safety injection signal is generated wnen the appropriate setpoint is reached.

These countermeasures will limit the consequences of the accident in two ways:

1.

Reactor trip and borated water injection supplement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

However, no credit is taken in the LOCA analysis for the boron content of the injection water.

In addition, the insertion of contr".'.

rods to shut down the reactor is neglected in the large break analysis.

2.

Injection of borated water provides for heat transfer rrom the core and prevents excessive clad temperatures.

Oescr ation of Large Break Loss-of-Coolant Accident Transient The sequence of events following a large break LOCA is p".esen ed in Table

14. 3. 1-6.

Before the break occurs, the unit is in an equilibrium condition; that is, the heat generated in the core is being removed via the secondary system.

Ouring blowdown, heat from fission product decay',

hot internals and the vessel, continues to be transferred to the reactor coolant.

At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling.

After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10 CFR 50.( 'hereafter the core heat transfer is (1)

unstable, with both nucleate boiling and film boiling occurring.

As the core becomes uncovered, both turbulent and laminar forced convection and radiation are considered as core heat transfer mechanisms.

The heat transfer between the RCS and the secondary system may be in either direction, depending on the relative temperatures.

In the case of continued heat addition to the secondary

system, the secondary system pressure increases and the main steam safety valves may actuate to limit the pressure.

makeup water to the secondary side is automatically provided by the emergency feedwater

.system.

The safety injection signal ac uates a feedwater isolation signal which isoiaies normal feedwater flow by closing the main feedwater isolation valves, and also initiates emergency feedwater flow by starting the emergency feedwater pumps.

The secondary flow aids in the reduction of RCS pressure.

'>)hen the RCS depressurizes to 600 psiz, the accumulators begin to inject borated water into the reactor coolant loops.

The conservative assumption is made that accumulator water injec ed bypasses the core and goes out through the break until the termination of bypass.

This conservatism is again consistent with Appendix K of 10CFRSO.

Since loss of offsite power (LOOP) is assumed, the RCPs are assumed to trip at the inception of'the accident.

'The e'ffects of'ump coastdown are i'nc'luded

'n the blowdown analysis.

The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2280 psia) falls to a value approaching that of the containment atmosphere.

Prior to or at the end of the blowdown, the

mechanisms that are responsible for the emergency core cooling water injected into the RCS bypassing the core are calculated not to be effective.

At this time (called end-of-bypass) refill of the reactor vessel lower plenum begins.

Refill is completed when emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods (called bottom of core recovery time).

The reflood phase of the transient is defined as the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated.

From the latter stage of blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer.

The downcomer wa'ter elevation head provides the driving force required for the reflooding of the reactor core.

The low head and high head safety injec ion pumps aid in the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the reflooding process.

Cont'inued operation of the ECCS pumps supplies wa er during longterm cooling.

Core temperatures have been reduced to longterm steady state levels associated with dissipation of residual heat generation.

After tne water level of the residual water s orage tank (RWST) reaches a

minim m allowable value, coolant for long-tern cooling of the core is obtained by switching to the cold recirculation phase of opera ion, in which spilled borated wa er is drawn from the engineered safety ea ures (ESF) containment sumps by the low head safety injection (residual heat removal) pumps and returned to the RCS cold legs.

The containment spray system continues to operate to further reduce containment pressure.

r r.

Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 'after initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control the boric acid concentra-ion in the reactor vessel.

Core and S stem performance Mathematical Model:

The requirements of an acceptable CCS evaluation model are presented in Appendix K of 10 CFR 50 (Federal Register 1974). (1)

Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases:

(1) blowdown, (2) refill, and (3) ref lood.

There are three distinct transients analyzed in each

phase, including the thermal-hydraulic transient in the

RCS, the pressure and temperature transient within the containment, and the fuel and clad temperature transient of the hottest fuel rod in the core.

Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.

A description of the various aspects of the LOCA analysis methodology is given by Bordelon, Massie, and 2ordan

( 1974).

Tnis document (6) describes the major pnenomena

modeled, the inter-.aces among the computer

codes, and the features of the codes which ensure comoliance with the Acceptance Criteria.

The SATAN-'ll, WREFLGO0, BART and LOCTA-IV codes, wnich are used in the LOCA analysis, are described in detail by Bordelon et al. (1974)( 'elly et al.

(1974)

'; Young et al.

(5)

(9)

(4)

(1980);

Bordelon and Murphy (1974)( ';

and Bordelon et al.

/~X

( 1974).

Code modifications are specified in References 2,

7 and 13.

These codes assess the core heat transfer geometry and determine if the core remains amenable to cooling throughout.and subsequent to the blowdown, refill, and reflood phases of the LOCA.

The SATAN-V1 computer,

~,

' co'de'nalyzes the thermal-hydraul'ic;'transi'ent in 'the RCS during blowdow'n and the WREFLOOO computer code calculates thi's transient during the refill and reflood phases to the accident.

The LOTiC computer

code, described by Hsieh and Raymund in

0

WCAP-8355 ( 1975) and WCAP-8345 ( 1974)

, calculates the containment (3) pressure transient.

The containment pressure transient is input to WREFLOOO for the purpose of calculating the reflood transient.

The LOCTA-IV computer code calculates the thermal transient of the hottest fuel rod during the three phases.

The standard Pad Fuel Thermal Safety Model, described in Reference 15, generates the initial fuel rod conditions input to LOCTA-IV.

SATAN-VI calculates the RCS pressure, enthalpy, density, and the mass and energy flow rates in the

RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA.

SATAN-VI also calculates the accumulator water mass and internal pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown.

At the end of the blowdown phase, these data are transferred to the WREFLOOD code.

Also, at the end-of-blowdown, the mass and energy release rates during blowdown are input to the LOTIC code for use in the determination of the contai.",".ien pressure response during <<his first phase of the LOCA.

Additiona',

SATAN-VI outout data from the end-of-blowdown, including the core inlet flow rate and

enthalpy, the core pressure, and the core power decay transi'ent, are input to the LOCTA-IV code.

With input from the SATAN-VI code, WREFLOOO uses a system thermal-hydraulic model to determine the core flooding rate (".hat is, he rate at wnich coolant enters the bottom of'he core),

he cool-ant pressure and"temperature, and the quench front height during the reflood phase of the LOCA.

WREFLOOO also calculates the mass and energy flow addition to the containment through the break.

WREFLOOO is also linked to the BART and LOCTA-IV codes.

The heat transfer calculation for the I

7 II

, average fuel channel in the hot assembly during the ref lood phase of the

~

LOCA is performed by the BART'omputer code using a mechanistic (16) core heat transfer model.

This information is then used by LOCTA-IV throughout the analysis of the LOCA transient to calculate the fuel clad temperature and metal-water reaction of the hottest rod in the core.

The large break analysis was performed with the December 1981 version of (16) the Evaluation Model modified to incorporate the BART computer code.

Input Parameters and Initial Conditions:

The analysis presented in, this section was performed with a reactor vessel upper head temperature equal to the RCS hot leg temperature.

The bases used o select the numerical values that are input parameters to the analysis have been conservatively determined from extensive sensitivity studies (Westinghouse 1974

Salvatori 1974 (12).

~

(11).

Johnson, Hassle, and Thompson 1975

).

In addition, the requirements (8) of Appendix K regarding specific model 'features were met by selecting models which provide a significant overall conservatism in the analysis.

The assumptions which were made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs, and include such items as the core oeaking factors, the containment

pressure, and the performance or the

=CCS.

Gecay heat generated throughout the transient is also conservatively calculated.

A meeting was held at :he Mestinghouse Licensing Office in Bethesda on Oecember 17, 1981 between members of the U.

S. Nuclear Regulatory Commission and members of tne Mestinghouse Nuclear 'Safety Oepartment to discuss the impac. of maximum safety injection on the large break ECCS analysis on a generic basis.

Further discussion of this issue is provided in a letter from E.

P.

Rahe, Manager of Westinghouse Nuclear Safety Oepartment, to Robert L. Tedesco of the U.

S. Nuclear Regulatory Commission.(,

A brief description of this issue is given below.

(14)

.,Mestihghause ECCS analyses currently: assume. minimum s'afeguards for the

~

safety injection flow, which minimizes the amount of flow to the RCS by

~

assuming maximum injection line resistances, degraded ECCS pump performance, and the loss of one residual heat removal (QHR) pump as the most limiting single failure.

This is the limiting single failure assumption when offsite power is unavailable for most Westinghouse

~

~

plants.

However, for some Westinghouse plants including 0.

C.

Cook Unit 1, the current nature of the Appendix K ECCS evaluation models is such that it may be more limiting to assume the maximum possible ECCS flow delivery.

In that case, maximum safeguards which assume minimum injection line resistances, enhanced ECCS pump performance, and no single failure, result in the highest amount of flow delivered to the RCS.

Current LOCA analysis for 'the 0.

C.

Cook Unit 1 has demonstrated

that, maximum safeguards assumptions result in he highest peak clad temperatur'e.

Therefore, the worst break for O.

C.

Cook (CO = 0.6) was re-analyzed, assuming maximum safeguards.

Results:

Based on the results of the LOCA sensitivity studies (Westinghouse 1974

Salvatori 1974

Johnson,

Massie, and Thompson 1975

) the limiting large break was found to be the double ended cold leg guillotine (OECLG).

Therefore, only the OECLG break is considered in the large break ECCS performance analysis.

Calculations were performed for a range of Moody break discharge coefficients.

The results of these calculations are summarized in Tables 14.3. 1-5 and 14.3.1-6.

The containment data used to generate the LOTIC backpressure transient are shown in Table 14.3. 1-1.

The mass and energy release data ror the minimum and maximum safeguards cases are shown in Tables 14.3. 1-2 and 14.3. 1-3 respectively.

Nitrogen release rates to the containment are given in Table 14.3.1-4.

Figures 14.3. 1-1 through 14:3. 1-54.present the transients for the

'I principal parameters 'for the break size's analyzed.

The following items are noted:

Fi ures 14.3.1-1 throu h 14.3.1-12 The following quantities are presented at the clad burst location and at the hot spot (location of maximum clad temperature),

both on the hottest fuel rod (hot rod):

1.

fluid quality; 2.

mass velocity; 3.

heat transfer coefficient.

The heat transfer coefficient shown is calculated by the LOCTA-IV code.

Fi ures 14.3.1-13 throu h 14. 3. 1-24 The system pressure shown is the calculated pressure in the core.

The flow rate from the break is plotted as the sum of both ends for the guillotine break cases.

The core pressure drop shown is from the lower plenum, near the core, to the upper plenum at the core outlet.

Figures 14.3.1-25 throu h 14. 3. 1-36 These figures show the hot spot clad temperature transient and the clad temp rature transient at the burst location.

The fluid emperature shown is also for the hot spot and burst location.

The core flow (top and bottom) is also snown.

Figures 14.3. 1-37 These figures show he core rerlood transient.

through 14.3. 1-44 Figures 14.3. 1-45 throu h 14. 3. 1-52 These figures show the mergency Core Cooling System flow for all of the cases analyzed.

As described earlier, the accumulator delivery during blowdown is discarded until the.end of bypass is calculated.

Accumulator 'flow:, however, is established in the refill and the reflood calculations.

The accumulator flow assumed is the sum of that injected in the intact cold legs.

Fi ures 14.3. 1-53 throu h 14.3.1"54 The containment pressure transient used in the analysis is also provided for the minimum and maximum SI cases.

Figures 14.3.1-55 and 14.3.1-60 These figures show the heat removal rates of the heat sinks found in the lower compartment and the heat removal by the lower containment drain, andthe heat removal by the sump and LC sprays (minimum and maximum SI cases).

Fi ures 14.3.1-61 throu h 14. 3. 1-64 These figures show the temperature transients in both the upper and lower compartments of the containment and flow from the upper to lower compartments.

Total heat removal in the lower compartment is the sum of all the heat removal rates shown (for minimum and maximum SI cases).

The maximum clad temperature calculated for a large break is 2163 F,

which is less than the Acceptance Criteria limit of 2200~F.

The maximum local metal-water reaction is 9.65,percent, which is well below the embrittlement limit.of 17 percent as required by 10 CFR 50.46.

The

otal core metal-water reaction is less than 0.3 percent for all breaks, as compared with the 1 percent criterion of 10 CFR 50.46.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.

As a result, the core temperature will continue o drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.

10

References for Section 14.3. 1. 1 1.

"Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors,"

10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Re ister

1974, Volume 39, Number 3.

2.

Rahe, E.

P. (Westinghouse),

letter to J.

R. Miller (USNRC); Letter No. NS-EPRS-2679, November 1982.

3.

Hsieh, T.,

and

Raymund, M., "Long Term Ice Condenser Containment LOTIC Code Supplement 1," WCAP-8355, Supplement 1,

May 1975, WCAP-8345 (Proprietary), July 1974.

4.

Bordelon, F.

M. et "al., "LOCTA-IV Program:

Loss-of-Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-proprietary),

1974.

5.

Bordelon, F.

M. et al.,

"SATAN-VI Program:

Comprehensive

Space, Time Oependent Analysis of Loss-of-Coolant,"

WCAP-8302 (Proprietary) and WCAP-8306 (Non-proprietary),

1974.

6.

Bordelon, F. M.; Massie, H. W.; and 2ordan, T. A., "Westinghouse ECCS Evaluation Model - Summa'ry,"

WCAP-8339, 1974.

7.

Rahe, E. P.,

"Westinghouse ECCS Evaluation Model, 1981 Version,"

WCAP-9220-P-A (Proprietary Version),

WCAP-9221-?-A (Non-proprie~ry version),. Revision 1,

1981.

8.

Johnson, W. J.; Massie, H. W.; and

Thompson, C. M., "Westinghouse ECCS - Four Loop Plant (17x17) Sensitivity Studies,"

WCAP-8565-P-A (Propr'ietary) and WCAP-8566-A (Non-proprie't'ary),

1975.

PP 9.

Kelly, R. 0. et al., "Calculational Model for Core Ref looding After" a Lo'ss-of-Coolant Accident (WREFLOOO Code),"

WCAP-8170 (Proprietary) and WCAP-8171 (Non-proprietary),

1974.

10 U. S. Nuclear Regulatory Commission 1975 "Reactor Safety Study - An Assessment of Accident Risks in U.

S.

Commercial Nuclear Power Plants,"

WASH-1400, NUREG-75/014.

11. Salvatori, R., "Westinghouse ECCS - Plant Sensitivity Studies,"

WCAP-8340 (Proprietary) and WCAP-8356 (Non-proprietary),

1974.

12.

"Westinghouse ECCS - Evaluation Model Sensitivity Studies,"

WCAP-8341 (Proprietary) and WCAP-8342 (Non-proprietary),

1974.

13. Bordelon, F. H., et al.,

"Westinghouse ECCS Evaluation Model-Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-proprietary),

1975; 14.

Rahe, E.

P. (Westinghouse).

Letter to Robert L. Tedesco (USNRC),

Letter No. NS-EPR-2538, Oecember 1981.

15. Letter from J.

F. Stoltz (NRC) to T.

M. Anderson (Westinghouse);

subject:

Review of WCAP-8720, Improved Ana'.ytical 4lodels Used in Westinghouse Fuel Rod Oesign Computations.

16.

Young, 4I. Y., et al.,

"BART-Al:

A Computer Code for he Best Estimate Analysis of Reflood Transients, "WCAP-9561-P-A (Proorietary) and WCAP-9695"A (Non-Proprietary)

January 1980.

12

TABLE 14.3

~ 1-1 LARGE BREAK CONTAINMENT DATA

( ICE CONDENSER CONTAINMENT)

NET FREE VOLUME

( Includes Distribution Between

Upper, Lower, and Dead-Ended Compartments)

UC LC OE IC 746,829'ft.

249,446 116,168 122,400 Initial Conditions Pressure Temperature for the Upper, Lower and Dead-Ended Compartments RWST Temperature Service Mater Temperature Temperature Outside Containment Initial Spray Temperature 14.7 psia UC 100~ F LC 120~ F OE 120oF 70~F 40oF 7oF 70~F Spray System Burnout Flow for a Spray Pump Number of Spray Pumps Operating Post-Accident Initiation of Spray System Ois ribution of the Spray Flow to the Upper and Lower Compartments 3600 gpm 2

40 secs LC 2835 gpm UC 43o5 gpm Deck Fan Post-Accident Initiation of Deck Fans Flow'at,e Per Fan 600 secs

~ 39,000 cfm per'ran

'I Hydrogen Skimmer System Flow Rate 2800 cfm per ran Assumed Spray Efficiency of Mater from Ice Condenser 'Drains 100'o 13

TABLE 14. 3 1 ~ 1

'continued)

STRUCTURAL HEAT SENKS i

2 blateri a 1 1.

LC 2.

LC 3.

LC 4.

LC 5.

LC 6.

LC 7.

LC 8.

LC 9.

LC 10.

LC 11.

LC 12.

LC 13.

UC 14.

UC 15.

UC 16.

UC 17.

UC 18.

UC 19.

UC 12,105 11,700 65,980 5,481 4,735 289 14,690 3,439 5,775 4,966 7,013 2,457 378 29,772 8,033 420 29,330 34 125 2'10 0.0469/2.0 2.0 1.35 0.0833 0.01147 0.25 0.0079 0.1561 0.009 0.0096 0.037 0.0334

.1667/.0365

.0092

.0209

.0052 1.47 0.0469/2.0

.0052 steel/concrete concrete concrete steel steel lead steel steel steel steel steel steel steel/concrete steel steel s eel concrete steel/concrete steel UC:

Upper Compartment

~:......

LO:..Lower. Compartment OE:

Oead-Ended Compartment 1C:

lce Condenser Compartment

ASS AiND 9

?"=Y RE':-~SE ?~iES u!,.<IMUM Si i !!ME (sec)

O.

.2GGGE.Q1

.cuGQEiyl

.6GGGE'v I

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, 1 ZC'C E '02

.124ciE 02

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~ 15GGE+GZ

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.200GE'02

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.220GE 02

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TABLE 14.3.1-A-3 NITROGEN MASS AND ENERGY RELEASE RATES

~Time sec Flow Rate lbs/sec 37.5 39.5 45.5 47.5 53.5 55.5 61.5 63.5 70.3 72.3 78.5 80.5 86.3 88.3 94.3 96.3 102.2 104.2 110.2 112.2 126.2 128.2 138.2 140.2

~

146.2 148.2 174.2 176.2 71.9'0.7 37.2 31.6 18.8 15.6 8.5 6'. 9 186.0 158.0 97.3 82.4 48.5 40.0 21.9 18.2 11.7 10.5 7.6 6.8 3.3 2.9 1.8 1.6 1.2'.1 0.25 0.075 0329L:6/840727 l7

TABLE 14.3.1-A-4 LARGE BREAK OECLG C0=0.4 Results Max SI Peak Clad Temp. 'F Peak Clad Location Ft.

Local Zr/H20 Reaction (Max)i~

Local 2r/H20 Location Ft.

Total 2r/H20 Reaction,'o'ot Rod Burst Time sec.

Hot Rod Burst Location Ft.

2162 7,. 50 6.58 7.50

( 0.3 71.4 6.75 Calculation Licensed Core Power (Mwt) 102;o'f Peak Linear Power (kw/ft) 102;o'f Peaking Factor (at License Rating)

Accumulator Water Volume (ft ) per Accumulator 3

3250 13.225 1.97 950

. Cycle Analyzed.,:Cycle 8

18 0329L:6/840727

TABLE 14.3. 1-A-5 LARGE BREAK TIME SEQUENCE OF EVENTS Max SI OECLG CO=0.4 (sec)

START Reactor Trip Signal Safety Injection Signal Accumulator Injection End of Blowdown Bottom of Core Recovery Accumula'tor Empty Pump Injection 0.00 0.60 4.05 20'. 50 38.70 52.78 67.45 29.05

~

~

0329L: 6/840727 19

TABLE 14. 3.1-4 NITROGEN MASS AND ENERGY RELEASE RATES Time sec Flow Rate lbs/sec 37.5 39.5 45.5 47.5 53.5 55.5 57.5 60.7 66.7 68.7 74.7 76.7 78.7 80.7 86.7 88.7 98.7 100.7 110.7 112.7 122.7 124.7 130.7 132.7

'146. 6 145.5 71.9 60.7 37.2 31.6 18.8 15.6 12.8 266.81 159.7 135.7 83.2 70.3 58.9 49.1 27.2 22.3 10.7 9.6 5.6 5.1 3.0 2.7 2.0 1.8

~ 0,8 0.7 0329L:6/840727 20

TABLE 14.3.1-5 LARGE BREAK Results OECLG C =0.8 D

Min SI DECLG CD=0.6 Min SI OECLG

~

C =0.4 0

Min SI OECLG

'O='6 Max SI Peak Clad Temp.,

F 1942 Peak Clad Location, ft 7.00 Local Zr/H20 Reaction (Max) 2.85 Local Zr/H20 Location, ft 7.00 Total Zr/H20 Reaction

<0.3 Hot Rod Burst Time, sec 43.8 Hot Rod Burst Location, ft 6.00 2014 5.75 5.65 5.75

<0.3 37.8 5.75 1956 7.00 3.84 5.75

<0.3 47.4 5.75 2163 6.00 9.65 5.75

<0.3 37.8 5.75 Calculation Licensed Core Power (MWT) 102;< of 3411 Peak Linear Power (kw/ft) 102;~ of 14.796 Peaking Factor (at License Rating) 2.10 Accumulator Water Volume (ft ) per Accumulator 950 3

21

~ 0329L:6/840727

TABLE 14.3. 1-6 LARGE BREAK TIME SEQUENCE OF EVENTS Min SI OECLG CO=0.8 (sec)

Min SI OECLG CO=0.6 (sec)

Min SI OECLG CO=0.4 (sec)

Max SI OECLG CO=0.6 (sec)

START Reactor Trip Signal Safety Injection Signal Accumulator Injection End of Blowdown Bottom of Core Recovery Accumulator Empty Pump Injection 0.00 0.00 0.62 0.63 3.83.

3.95 12.90 15.50 29.68 30.43 40.66 43.29 56.89 59.29 28.83 28.95 0.00 0.64 4.20 20.80 38.49 52.64 65.65 29.20 0.00

0. 63.

3.95 15.50 30.43 42.47 60.58 28.95 0329L: 6/840727 22

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ATTACHMENT 8 LOCA Rci A-"0 -."CH SPECS Plant Name:

Oonald C.

Cook Unit 1

(AFP)

Tyoe/Date o

'OCA Analysis O.lp 0C

= ~-4 (max SI case)

Large Break LOCA Analysis Df'Qg Total Peaking Factor F~. ~ g,'IQ Cold Leg Accumula or.

Water Yolume:

980 ft /accumulator (nominal) unchanged

-,rom cur rent tech specs Cold Leg Accumulator Gas Pressure:

600 psia (minimum) unchanged from current tech specs K(z) Cu. ve:

See next page

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