ML20114A971

FSAR Chapter 15 Sys Transient Analysis Methodology

ML20114A971
Person / Time
Site:	Catawba, McGuire, Mcguire
Issue date:	11/30/1991
From:	DUKE POWER CO.
To:
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ML20114A969	List: ML20114A965 ML20114A971
References
DPC-NE-3002-A, NUDOCS 9208240197
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DUKE POWER COMPANY McGUIRE !JUCLEAR STATI0t1 i- CATAWBA !JUCLEAR STATION FSAR CHAPTER 15 SYSTD4 TRA!4SIEllT A!!ALYSIS METHODOLOGY DPC !JE-3002-A

!JOVEMBER 1991 s

b Nuclear Engineering Group Nuclear Services Division liuclear Generation Department ,

Duke Power company 9208240197 920917 PDR ADOCKc05000369- - -

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L DUKEPOWER August 30, 1993 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 _

subject: McGuire Nuclear Station Docket Numbers 50-369 and -370 Catawba Nuclear Station Docket Numbers 50-413 and -414 FSAR Transient Analysis Methodology; .

Topical Report DPC-NE-3002-P Attached for your review is Duke Power Company's Topical Report DPC-NE-3002, " TSAR Chapter 15 System Transient Analysis Methodology." This report describes Duke's methodology for conservatively modeling those FSAR Chapter 15 non-LOCA transients and accidents not previously described in DPC-NE-3000, " Thermal-Hydraulic Transient Analysis Methodology" and DPC-NE-3001,

" Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology." This report is applicable to the McGuire -

and Catawba Nuclear Stations.

The objectives of this report are: 1) to describe the initial and boundary conditions and input assumptions regarding control and protective system functions, as used in the analysis of FSAR Chapter 15 events; and 2) to describe nodalization and/or modeling dif ferences relative to those analyses previously detailed in DF C-NE-3000. The rod ejection, steam line break, and dropped rod methodologies are described in DPC-NE-3001, and are not discussed in this report. Assumptions regarding safety analysis physics parameters are also discussed in DPC-NE-3001.

Please note that approval of this Topical Report is needed for startup of McGuire U1.it 1 Cycle 8 following its upcoming refueling outage. The outage is scheduled to begin in lata September 1991.

Cycle 8 is expected to start up in late November or early D. ember.

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'j c4 U. S. Nuclear Regulatory Commission August 30, 1991 Page 2 If there are any questions, please call Scott Gewehr at (704) 373-7581.

Very truly y urs, Si W M. S. Tuckman cvr3002/ sag cc: Mr. T. A. Reed, Project Manager '

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 9H3, OWFN Washington, D. C. 20555 Mr. R.E. Martin, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 9H3, OWFN Washington, D. C. 20555 Mr. S. D. Ebneter, Regional Administrator' U.S. Nuclear Regulatory Commission - Region II _

101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 Mr. P. K. Van Doorn Senior Resident Inspector McGuire Nuclear Station Mr. R. C. Jones Reactor Systems Blanch Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station C _ . . .

/ 'o, UNITED STATES '

NUCLEAR REGULATORY COMMISSION 5 -t WA EHlNGTON, D. C. 20555 i

'% , , , , , / November 15. 1991 Docket Nos. 50-369, 50-370 50-413 and 50-414 Mr. N. B. Tucker, Senior Vice President Nuclear Generation

Duke Power Company P. O. Box 1007 Charlotte, North Carolina 28201-1007

Dear fir. Tucker:

! SUBiECT:

1 SAFETY EVALUATION ON TOPICAL REPORT DPC-NE-300?, "FSAR CHAPTER 15 i

l SYSTEli TRAf'SIENT ANALYSIS HETH000 LOGY," (TAC NO. 66850)

The NPC staff with the support of its contractor has reviewed Duke Power Company Topical Report DPC-NE-300?, "FSAR Chapter 15 System Transient Analysis

!!ethodoloay," dated August 30, 1991, as supplemented by letters dated October 16 and November 5, 1991. The staff has found the topical report to be acceptable subject to the conditions identified in section 4.0 of the

! attached Technical Evaluation Report as modified by Section 2.? of the l attached Safety Evaluation.

This concludes our review activities in response to your submittals regarding Topic 61 Report OPC-NE-3002.

Sincerely,

/

Timothy A. Reed, Pro,iect lianager Project Directorate 11-3 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation

Enclosures:

1. Safety Evaluation P. Technical Evaluation Report cc: See next page -

-J

Catawba Nuclear Station Duke Power Company McGuire Nuclear Station CC:

Mr. R. C. Futrell Mr. Alan R. Herdt, Chief Regulatory Compliance Manager Project Branch 43 Duke Power Company U.S. Nuclear Regulatory Connission Catawba Nuclear Site 101 Marietta Street, NW, Suite 2900 Clover, South Carolina 29710 Atlanta, Georgia 303?3 Mr. A.Y. Carr, Esq. North Carolina Electric Membership Duke Power Company Corp.

422 South Church Street k P.O. Box 27300 Charlotte, North Carolina 28242-0001 Raleigh, North Carolina a 611 J. Michael McGarry,111. Esq. Saluda River Electric Cooperative, Winston and Strawn Inc.

1400 L 5treet, N.W. P.O. Box 929 Washington, DC 20005- Laurens, South Carolina 29360 North Carolina MPA-1 Senior Resident inspector Suite 600 Route 2, Box 179N P.O. Box 29513 York, South Carolina 29745 Raleigh, North Carolina 27626-513 Regional Administrator, Region 11 Mr. Frank Modrak U.S. Nuclear Regulatory Commission Project Manager, Mid-South Area 101 Marietta Street, NW, Suite 2900 ESSD Projects _

Atlanta, Georgia 30323 Westinghouse Electric Corporation

.MNC West Tower - Bay 241 Pr. Heyward G. Shealy, Chief P.O. Box 355 Bureau of Radiological Health Pittsburgh, Pennsylvania 15230 South Carolina Dept. of Health and Environmental Control County Panager of York County ?600 Bull Street York County Courthouse Colunbia, South Carolina 29201 York, South Carolina 29745

' Ms. Karen E. Long t!ichard P. Wilson, Esq. Assistant Attorney General Assistant Attorney General North Carolina Dept of Justice S.C. Attorney General's Office P.O.-Box 629 P.O. Box 11549 Raleigh, North Carolina 27602 Colwhia, South Carolina 29211 Mr. R.'L. Gill, Jr.

Piedmont Municipal Power Agency Licensing 121 Village Drive Duke-Power Company Greer, South Carolina 29651 P.O. Box 1007 Charlotte, North Carolina 28201-1007 E . _ . . . .-. -

Catawba Nuclear Station Duke Power Company McGuire Nuclear Station County Manager of flecklenburg County Dr. JoFn M. Barry 720 East Fourth Street Department of Environmental Health Charlotte, North Carolina 28202 Mecklenburg County 1200 Blythe Boulevard Charlotte, North Carolina 28203 Mr. R. O. Sharpe Hr. Dayne H. Brown, Director Compliance Department of Environmental Health Duke Power Company and Natural Resources McGuire Nuc1 car Site Division of Radiation Protection 12700 Hagers Ferry Road P. O. Box 27687 Huntersville, North Carolina 20078-8985 Raleigh, North Carolina 27611-7607 ~

Senior Resident inspector fir. M. S. Tuckman c/o U.S. Nuclear Regulatory Comission Vice President, Catawba Site 12700 Hagers Ferry Road Duke Power Company Funtersville, North Carolina 28078 P. O. Box 256 Clover, Snuth Carolina 29710 Mr. T. C. Nctieekin Vice President, McGuire Site Duke Po.ier Company 12700 Il69trs Ferry Road Huntersville, North Carolina 28078-8905

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j ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT DPC-NE-3002

" TSAR CHAPTER 15 SYSTEM TRANSIENT ANALYSIS METHODOLOGY" O_ UKE POWER COMPANY h

MCGUIRE NUCLEAR STATION CATAWBA NUCLEAR STATION DOCKET NOS. 50-369, 50-370, 50-413 AND 50-414

1.0 INTRODUCTION

1 By letter dated August 30, 1991, the Duke Power Company (DPC) submitted Topical Report DPC-NE-3002, McGuire Nuclear Station and Catawba Nuclear Station, "FSAR Chapter 15 System Transient Analysis Methodology," describing modelling assumptions used by DPC in performing analyses of FSAR Chapter 15 events. This report, as supplemented by letters of October 16 and November 5, 1991, is intended to augment Topical Report DPC-NE-3000, "The Thermal-Hydraulic Transient Analysis Methodology - Oconee Nuclear Station, McGuire Nuclear Station, Catawba Nuclear Station." DPC-NE-3002 is also related to DPC-NE-2004, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01," and DPC-NE-3001,

" Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology."

2.0 STAFF EVALUATION The staff-performed its evaluation of the methodology reported in DPC-NE-3002 with i.he technical assistance of International Technical Services, Inc.

(ITS). The evaluation and findings are described in detail in the ITS technical evaluation report (TER) which is enclosed as part of this report.

As identified in the TER, certain items from OPC-NE-3002 were not included in

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this review because they have already been included in the review of one of the other related DPC topical reports. For instance, steam line break, control rod misoperation, and rod ejection events are included in the DPC-NE-3001 review and not repeated herein except as reference.

2.1- Other items Not Evaluated in TER L 2.1.1 Boron Dilution Event The TER identifies that the review of this event is beyond its scope.

I DPC-NE-3002 discusses boron dilution events. However, apart from core physics I aspects of DPC-NE-3001, the DPC methodology for evaluating boron dilution events does not use the codes described in the related topical reports identified in Section 1 of this SER. The staff concludes that the find.ng of acceptability for the boron dilution event analysis methodology of record continues to apply.-

2.1.2 Steam Generator Tube Rupture (SGTR)

The TER identifies that the review of this event is beyond its scope. DPC-NE-3002 discusses SGTR events; however, except for any parts of DPC-NE-3001 that may be found to apply, the DPC methodology for evaluating SGTRs does not use codes described in the related topical reports identified in Section 1 of this SER. The staff concludes that the finding of acceptability for the SGTR analysis methodology of record continues to apply.

l 2.2 TER' CONCLUSIONS i

- 2,2.1 Feedwater Line Break I'

TER Section-4.0 (Conclusions) recommends'that justifications for trip and

-- actuation times be required when the methodology is applied.

While the staff agrees that trip setpoints and actuation times must be

- consistent with the assumptions in FSAR analyses, we find that this I

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- 3 consistency is implemented in the plant technical specifications and is outside the scope of OPC-3002 and this re. view.

2.2.2 Power and Reactivity Feadback TER Section 4.0 recomends that the modelling of power and reactivity feedback be reviewea ano that it be assured that such modelling has n> adverse effect on the other modelling described in the TER. The staff review of * -

NE-3001 ;

covered these considerations and found them acceptable.

2.2.3 Locked Rotor Event TER Section 4.0 identifies that DPC has proposed that reactor coolant system (RCS) pressure of 120% of design pressure be used as a performance acceptance criterion for locked rotor event analyses replacing the previous 110%

criterion. Based on our review we find that the licensee has not provided adequate justification for the proposed change, particularly in light of the crer' ' taken in the DPC methodology for delayed loss of power to the unlocked reactor coolant pumps. The licensee identifies that its locked rotor event analyses calculate a peak RCS pressure of less than 110% design pressure. We find the DPC locked rotor analysis methodology (incorporating the 110% RCS pressure criterion) and results < ble.

2.2.4 Parametric Studies TER Section 4.0 recommends that parameteric studies be r.equired to be presented when the meth'odology is applied. The licensee has indicated that it will perform such studies, as needed. The staff finds this commitment acceptable.

3.0 STAFF CONCLUSIONS The staff finds the DPC transient analysis metnodology acceptable for McGuire and Catawba analyses.

Date: November 15, 1991 L

ENCLOSURE 2 ITS/NRC/91 10 L

TECHNICAL EVALUATION Qf_T_HE FSAR CHAPTER 15 SYSTEM TRANSIENT ANALYSIS METHODOLOGY TOPICAL REPORT DPC-NE-3002 FOR THE DUKE POWER COMPANY MCGUIRE AND CATAWBA NUCLEAR STATIONS I

1.0 INTRODUCTION

The topical report entitled "FSAR Chapter 15 System Transient Analysis Methodology," DPC-NE-3002, dated August 1991 (Ref.1), documents description of modeling assumptions used by Duke Power Company in performing transient analysis of FSAR Chaptt:r 15 accidents by discussing specific choices for L.e of the models described and qualified in DPC-NE-3000 using the RETRAN and VIPRE 01 computer codes (Refs. 2 and 3).

DPC documented, for licensing application, the conservative nature of (1) the

.. RETRAN model nodalization, (2) RdTRAN control systems, (3) use of the models described in the DPC-NE-3000 (Ref. 4) and (4) selection of initial and boundary conditions.

1.1 Scoce of Review Review of the subject topical report focused upon evaluation of acceptability, for licensing type analyses, of RETRAN models such as: (1) nodalizations for steam generators, core and reactor vessel, including any transient specific modifications; (2) selection of RETRAN internal models/ correlations and (3) selecticn of RETRAN initial and boundary condit'ons.

The topical report was further reviewed to assure that the application of DPC's DNB methodology was acceptable and consistent with the contents of DPC-NE-2004, DPC-NE-3000 and their supporting documents (Refs. 4 - 8) together 1

i with their respective TERs (Ref 9 and 10). The review, therefore, included identification of which transients DPC intends to analyze using its statistical core design (SCD) methodology and which they do not, and evaluation of DPC's selection of initial and boundary conditions in the systems analysis which was used to determined the statepoints for the DNB analysis.

Althour the subject topical report covered all applicable non-LOCA accident in Sections 15.1 through 15.6 of the FSAR, no review was conducted of the details of the transients which are presented in separate topical reports (steam line break, control rod misoperation, rod ejection and steam generator tube rupture) or those accidents identified by the DPC as: (i) not applicable to M/C plants; (ii) no system analysis deemed necessary; or (iii) those current licensing bases bounded by other analyses.

The following items are beyond the scope of this review: (i) review with -

respect to the core physics parameters or dose analyses; (ii) review related to the current McGuire 1 Cycle 8 (MIC8) reload analysis submittal; (iii) review of FSAR analyses; (iv) review of the Boron dilution event;(v) review of a statically misaligned control rod; and (vi) review of consistency or satisfaction of current Technical Specification., or proposed changes therein.

Therefore, no consistency check was made of DPC's philosophical approach documented in the topical report against the MIC8 reload analyses, FSAR analyses or Tect nical Specification limits. Furthermore, accuracy of details of the Reactor Protection System, Engineered Safety Features, instrumentation and auxiliary systems and their associ ted tolerance or uncertainty was not reviewed.

g Finally, no technical review was conducted as to the validity of DPC's assumption of 12K. of design pressure as an acceptance criterion for the RCP locked rotor analysis.

2.0 $EiARY Topical Report DPC-NE-3002 documents DPC's approach to performance of the 2

r _______ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - -_ - -- _. - - - -

I NSSS primary and secondary system analyses of FSAR Chapter 15 accidents. It covers all applicable non-LOCA accidents in Sections 15.1 through 15.6 of the FSAR except steam line break, dropped rod, and rod ejection, which are addressed in a separate topical report, DPC-NE-3001 (Ref,11).

DPC-NE-3002 presents brief discussion of specific choices for the use of the RETRAN plant models described in DPC-NE-3000, including nodalization, initial and boundary conditions ard modeling of the process instrumentation and control systems. Also presented are assumptions related to the Reactor Protection System, the Engineered Safety Features Actuation System, and _

av. of other systems and components. Trip actuation is discussed in oeno, .: 1 thus potential trip functions are presented. However, the wa r ' "

> no justification for actuation times for reactor trip, safety cther actions. Assumptions related to reactivity feedback er peaking and power distribution are not presented, therefore as viewed. Furthermore, although there is mention of intent to perform (or, in some instances, actual performance of) parametric studies to identify conservative scenarios and assumptions, none of such studies were presented.

The topical report contains qualitative, rather than quantitative information, and no the a:tual RETRAN or VIPRE computed results are presented. Therefore, this report presents DPC's philosophical approaches to 2 performance of FSAR Chapter 15 type analysis.

Nodalization selection is made based upon symmetry or a degree of asymmetry of the expected transient system response. Selection of initial and boundary conditions is designed to result in conservative predictions with respect to the aspect of a transient which the analysis is intended to assess, such as peak primary pressure, peak secondary pressure, short and long term core coolability. With respect to core coolability, selection of initial conditions depends upon the mode of DNBR computation; i.e., the use of the DPC developed SCD methodology SCD or the traditional DNBR methodology.

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3.0 E1A1UATI0l{

Acceptability of DPC's application of RETRAN models and assumptions for thermal-hydraulic calculations of FSAR Chapter 15 transient analysis of its McGuire/ Catawba (M/C) plants is discussed below. In addition, application to f

licensing type transient analysis of the SCD methodology dercribed in DPC Topical Report DPC NE-2004 and its supplements was also reviewed, i

i 3.1 McGuire and Catawba RETRAN_ Plant Model j The RETRAN base models for M/C plants were qualified in DPC-NE-3000 for both '

best-estimate and licensing type applications, subject to limitations {

described in the TER (Ref. 10).

DPC developed three different size models of the M/C Plants; a one-loop plant model to be used wher, all four loops are expected to behave similarly so that l there is no asymmetric condition; and a two-loop and a three-loop model to be used when more- detail is desirable due to asymmetric conditions expected in the reactor coolant system during the transient.

The steam generator model was examined in detail during review of DPC-NE-3000 for use in licensing analyses, specifically in over-pressurization transients. That review focused upon the ability of the DPC SG model to predict SG tube uncovery and resulting degradation of primary-to-secondary heat transfer. DPC presented results from an extensive sensitivity study to assure that during two transients considered, loss of normal feedwater and feedwater line break, the current modeling is adequate. The finding of that review is documented in the TER for DPC-NE-3000 and imposes certain

-limitations on use.

use of- certain RETRAN internal models such as the inter-region heat transfer model and local condition heat transfer model was reviewed and found to be acceptable for use in the components and for transients identified by DPC (Ref. 8).

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3.2 SCD Transients The core thermal-hydraulics for most of the transients considered in this topical report is analyzed using the OPC developed SCD methodology. For these transients, certain initial conditions used in the transient safety analysis are selected to be at nominal conditions, as qualitatively defined in Reference 10, since the uncertainty associated with the initial conditions is accounted for in the SCD method. These parameters are: (1) power level, (2) Core flow (RCS flowrate and core bypass flow), (3) Coolant temperature, -

and (4) RCS pressure. Other parameters necessary for the SCO method are not discussed in this topical report.

Those transient for which DNB is relevant but for which the SCD is not used are; (1) turbine trip. (2) RCP Locked Rotor, (3) startup of an inactive reactor coolant pump at an incorrect temperature, and (4) steam line break.

The turbine trip is not analyzed because as postulated, this transient results in a monotonically increasing DNBR which therefore is not an issue.

The SCD method is not used for DNBR analysis of steam line break since the primary pressure predicted during the transient is below the range of applicability of the CHF correlation used to develop the response surface equation. Similarly, the other events are outside the range of applicability -

of the response surface equation. -

3.3 Transient Initial Conditions and Assumotions In this section, initial and boundary conditions such as the transient initiators, reactor coolant pump operation and assumptions related to safety and relief valves are discussed. Control, protection and safeguard :ystem modeling is discussed highlighting which systems are credited or not credited, actuation logic and modeling assumptions.

A summary of assumptions and conditions selected by DPC is shown in Table 8.1 of the topical report as corrected by Reference 8. Definition of the terms used in the table are provided in Reference 8.

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Deviations from the following common analytical approach are highlighted in the ensuing sections of this TER:

1. For DNB analysis of SCD transients, SCD parameters are set at nominal while non-SCD parameters are set at con. frvative values.

2, For DNB analysis of non-SCD transients, all key parameters were set at conservative values.

3. For all DNB analyses except those which were initiated by reactivity insertion, the gap conductivity is assumed to be low to maximize the stored energy in the fuel and thereby minimize the change in heat flux out of the fuel during the transient, whereas for the reactivity insertion driven transients, the gap conductivity is assumed to be high because the transient duration is short compared to the fuel's thennal constant. For DNB analysis of transients which depressurize the primary, the pressurizer level is assumed to be at its high limit to maximize the depressurization.

4 Where transients are being analyzed for peak RCS pressure, the primary-to-secondary heat transfer is minimized, the pressurizer is assumed to be initi 'ly at the high limit of its operating range to produce the maximum pressure as the vapor region is compressed, and the fuel is assumed to have a high gap conductivity (which is accompanied by a low average fuel temperature) to maximize the energy transferred into the primary fluid.

5. For transients initiated on the primary side which have short duration, it is assumed that the results are insensitive to modeling of the secondary side and primary-to-secondary heat transfer. Therefore, for all such analyses the recondary side and steam generator parameters were set at nominal rather than conservative conditions.

6. Transients with symmetric loop behavior are analyzed with a single loop 6

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plant model while asymmetric transients are analyzed with a two loop model.

7. DPC uses the setpoint values and response time of trip function as specified in the Technical Specifications and accounts for uncertainty.

8. Decay heat is computed using the end-of-cycle data based .i ANSI /ANS-5.1-1979 standards plus a two-sigma uncertainty.

9. Availability assumptions on the PZR pressure and level con'.rol M mechanisms, such as the PZR sprays, PORVs and heaters, and the modes of operation are made in various combination to yield system behavior consistent with the transient being modeled. Steam line PORVs and condenser dump modeling is similar.

3.3.1 Increase in Heat Removal by the Secondary System Four transients are considered in this category; (1) feedwater (FW) system malfunctions that result in a reduction in feedwater temperature, (2) feedwater system malfunction causing an increase in feedwater flow, (3) excessive increase in secondary steam flow, and (4) inadvertent opening of a steam generator relief or safety valve. As stated earlier review of the -

steam line break event is beyond of the scope of this review. ~

The FW temperature reduction event is bounded by the FW flow increase event, which is analyzed. Since inadvertent opening cf a SG relief or safety valve is similar to, and bounded by, the steam line break, it is not analyzed, however a small step increase equC to 10?. of licensed core thermal power is presented in the report. Both of these transients are analyzed with respect to DNB using the SCD method.

An additional condition to consider a FW malfunction affecting more than one loop was recently added to the scenario of FW system malfunction event. DPC felt that the most limiting case would involve multi-loop malfunction affecting all loops eaually. Therefore, the use of a single-loop model is 7

appropriate.

The pressurizer liquid level is assumed to be high to maximize the primary presstre decrease. The SG mixture level is assumed to be low for the feedwater flow increase malfunction in order to maximize the overcooling before a protection or safeguards actuation. The small step increase in the steam flow event is not considered to be sensitive to SG level.

A conservatively large step change in main feedwater flow is assumad for the FW malfunction event. A 107. step increase in steam flow is as: for the other event.

In both event analyses, two cases are investigated to assess whether modeling the rod control system in manual control or automatic control would result in the worst case. In addition, minimum AFW flow, turbine trip ar.. FW isolation are credited and expected to trip on SG narrow range level after the appropriate Technical Specification response time delay.

The input selection and transient assumptions as described in the topical report for this category of events is acceptable.

3.3.2 Decrease in Heat Removal by the Secondarv System Four transient analyses are performed in this category: (1) turbine trip, (2) loss of offsite power, (3) loss of normal feedwater, and (4) feedwater system pipe break. Turbine trip is analyzed with respect to peak RCS and secondary side pressure, and the others are analyzed with respect to peak RCS pressure and DNB and/or long term core coolability (potential for hot leg boiling).

3.3.2.1 Turbine Trio DNBR analysis is not performed for this transient since this is a rapid transient in which prior to reactor trip, a significant RCS pressurization takes piace due to the reduction in secondary heat sink offsetting the increase in core inlet temperature, while the core power and the core flow 8

change very little. Therefore this event does not challenge the DNBR safety margin.

In peak RCS pressure analysis, reactor trip is expected to actuate on either overtemperature delta T (OTOT), overpower delta T (0PDT), or PZR high pressure. MFW is isolated upon turbine trip.

In the peak SG secondary side pressure analysis, RCS flow is assumed to be high to myimize the primary-to-secondary heat transfer. High SG level is assumed, to maximize the secondary pressure. In order to prevent a high PZR pressure reactor trip prior to OTDT trip, PZR PORVs are assumed operable.

3.3.2.2. Loss of Offsite Power This transient has potential challenges to peak RCS pressure, peak secondary side pressure, and DNB. However, the DNBR results from this event are bounded by the loss of flow event because these two events, as postulated by DPC, differ only in the timing of the insertion of the control rods. In the loss of offsite power (LOOP) event, the rods begin to fall immediately, whereas in the loss of flow event rods fall after an instrumentation delay.

Similarly, the peak primary system pressure is bounded by the loss of flow event. The secondary side pressure is bounded by the turbine trip event.

For LOOP, the reactor trips prior to the turbine trip, therefore by the time '-

the secondary pressure begins to increase, the primary system is rapidly cooling down. However, in the turbine trip event, reactor trip is after the turbine trip.

Therefore, a quantitative analysis of this transient is not required.

Nevertheless, DPC provided the analytical methodology for analysis of this event should it become necessary.

The transient will be analyzed with respect to three different objectives:

peak RCS pressure; peak secondary side pressure; and DNB using the SCD method.

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For peak RCS pressure analysis, all RCPs are tripped as the t ransie.nt initiating event. Reactor trip and MFW trip are assumed on LOOP. AFW is assumed to actuate on LOOP after a delay. However, in order to minimize the heat removal capability, the minimum AFW ficw is assumed.

For peak SG secondary side pressure analysis, OPC assumes high RCS flow to maximize the primary-to-secondary heat transfer. High SG level is assumed, to maximize the secondary pressure.

In order to determine statepoints to be used in DNB analysis using the SCD method, PZR level is assumed to be low to minimize the primary pressure ,

increase. Low SG level is assumed, which minimizes primary-te-secondary heat transfer.

3.3.2.3 Loss of Normal Feedwater The loss of normal feedwater is bounded by the turbine trip transient. Tht:

power to heat sink mismatch is greater for the turbine trip because the reactor trip and turbine trip occur simultaneously for the loss of FW event, while for the turbine trip event, reactor trip occurs after the turbine trip.

Therefore, a quantitative analysis of this transient is not required.

Nevertheless, DPC provided the analytical methodalogy for analysis of this event should it become necessary.

For peak RCS pressure analysis, reactor trip is assumed on the SG low-low level. AFV is assumed to actuate on the SG low-low level; however, in order to minimize the heat removal capability, the minimum AFW flow is assumed.

In order to maximize the peak SG secondary side pressure by maximizing the primary-to-secondary heat transfer, high RCS flow is assumed. High SG level is assumed, to maximize. the secondary pressure. Reactor trip is assumed on the SG low-low level. AFW is assumed to actuated on the SG low-low level with a minimum flow delivery.

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In order to determine statepoints to be used in DNB analysis using the SCD method, PZR level is assumed to be low to minimize the primary pressure increase. High SG level is assumed to delay reactor trip on SG low-low level. Reactor trip is assumed on the SG low-low level. AFW is assumed to actuate on SG low-low level with a minimum fiow delivery. Turbine trip is assumed on reactor trip.

3.3.C.4 Feedwater System Pine Break This transient is analyzed with respect to (1) DNB using the SCD method, and (2) long term core coclability (potential for boiling in the hot leg). The most limiting event assumed by DPC is the double-ended rupture of the largest feedwater line.

The DNB analysis for this transient is analyzed as a complete loss of coolant flow event initiated from an off-normal conditions. It is postulated in this transient that coincident with reactor trip (and turbine trip) loss of offsite power is assumed to occur causing RCP coastdown. Reactor trip is assumed on the OTDT. AFW is assumed to actuate on SG low-low level after a delay with a rainimum flow delivery in order to minimize the heat removal capability. Turbine trip is assumed on reactor trip.

Long Term Core Coolability (Hot Leg Boiling) -

A three-loop model is used since uneven flow of AFW into the unaffected SGs causes asymmetric loop behavior.

High core power is assumed to maximize the heat fl ux. PZR pressure is assumei to be low, which minimize:, the margin to hot leg boiling by lowering the hot leg saturation temperature. A high RCS temperature is assumed, to increase the amount of energy to be removed. Low SG level is assumed to maximize the loss of secondary heat sink. A high fuel temperature is assumed, accompanied by low gap conductivity. High SG tube plugging is assumed to minimize the primary-to-secondary heat transfer, 11

___ _____ _______-_____-__ ___-_____-__--_____-_-- - - a

The RCPs are assumed to trip at 15 seconds, which is assumed to precede the time at which the pumps would be manually tripped on high-high containment pressure.

Reactor trip is assumed at 10 seconds into the transient which is after the SI actuation on high containment pressure. SI actuation is assumed on high containment pressure at 10 seconds and terminated at 70 seconds when the emergency procedure criteria for termination are assumed to be met. AFW is l

assumed to actuate on 51 actuation after a delay. However, in order to  !

minimize the heat removal capability, the minimum AFW flow is a :ed . AFW <

1 is terminated at 120 seconds into the transient. MSIV closure are actuated l at IS seconds and assumed to precede automatic closure on high-high t containment pressura. Early closure is conservative in order to initiate the overheating portion of the transient. However, no justification was i presented for any of the actuation time assumptions.

The input selection and transient assumptions as de"ribed in the topical report. for this category of events is acceptable; however, trip actuation  ;

times must be justified in any application of this methodology. '

3.3.3 Decrease in Reactor Coolant System Flow Rate Three transient analyzed in this category are: (1) partial loss of forced reactor coolant flow, (2) complete loss of forced reactor coolant flow and l

(3) reactor coolant pump locked rotor. I 3.3.3.1 Loss of Forced RC Flow: Partial and Comolete Due to the similarity of these events, the partial loss of forced flow and l complete loss of forced flow events are discussed together.

A single-loop model is used for analysis of the complete loss of forced flow since the transient impacts all loops symmetrically: the two-loop model is used for the partial loss of forced flow event analysis. In both cases, DNB analysis will be performed using the SCD method.

12

For the partial loss of flow, a single reactor coolant pump is assumed to trip, while the other three pumps remain operational for the duration of the

( transient. For the complete loss of forced flow, all four RCPs are tripped at the initiation of the transient. The pump model is adjusted to yield pump coastdown which is conservative with respect to the flow coastdown test data.

Reactor trip for the partial loss event is assumed on low RCS flow after an appropriate delay time, while for the complete loss event, reactor trip is assumed on RCP undervoltage. Turbine trip is assumed on reactor trip.

3.3.3.2 RC Pumo locked Rctor This transient is analyzed with respect to both peak RCS pressure and DNB.

For both analyses a two-loop model is used for analysis due to the asymmetric nature of the transient.

In presenting its approach to these transients, DPC stated that it used an acceptance criterion of 120", design pressure. Review of this criterion is beyond the scc. e of this review.

In order to maximize RCS pressure, the RCS flow is assumed at its low initial flow to mininize the heat transfer to the secondary side. A high core bypa:,s J flow is assumed to minimize the core flow to max mize the heat-up. The initial RCS aver ge temperature is also assumed at its high level.

The transient initiating event is seizure of the rotor of the RCP in the faulted loop, while the other three pumps trip on bus undervoltage following the loss of cffsite power. Offsite power is assumed to be lost coincident with the turbine trip. Reactor trip is assumed on low RCS flow in the affected loop. Turbine trip is assumed on reactor trip. -

DNB analysis is performed using the traditional method. Therefore, core power is assumed to be high, while the PZR pressure and level are assumed to be low to minimize the pressure increase. High initial loop average 13

- -- _ _ _ _ - --------_---------------------D

temperature is assumed to maximize the stored energy in the primary which must be removed. Similarly, a high core bypass flow resulting in low core flow is assumed to maximize the heat-up and low RCS flow is chosen to maximize the primary-to-secondary heat transfer.

Offsite power is as::umed to be lost coincident with the turbine trip.

Similar to the peak RCS pressure case, reactor trip is assumed on low RCS flow in the affected loop. Turbine trip is assumed on reactor trip.

I The input selection and transient assumptions as described in the topical l report for this category of events is acceptable; however, the assumption .

that 1207. of design pressure is an acceptable limit must be reviewed by the  !

NRC staff.

{

3.3.4 Reactivity and Power Distribution Anomalies Seven transients are considered i '1 this category; (1) uncontrolled bhak 't I

withdrawal from a subcritical or low power startup condition, i (2) l uncontrolled bank withdrawal at power, _(3) statically misaligned control rod l (4) single control rod withdrawal, (5) startup of an inactive reactor coolant I pump at an incorrect temperature, (6) CVCS malfunction (boron dilution), and (7) inadvertent loading and operation of a fuel assembly in an improper l position. I Review of boron dilution event analysis and of inadvertent loading and operation of a fuel assembly in an improper position is beyond the scope of this review. Acceptability of these events should be rtviewed by an appropriate branch of HRC.

1 Each of the two uncontrolled bank withdrawal events is analyzed with respect to both peak RCS pressure and DNB. The single control rod withdrawal and startup. of an inactive RCP at in incorrect temperature are analyzed for DNB '

only. All transients except the . trtup of an inactive RCP are SCD transients.

14 I

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3.3.4.1 Uncontrolled Bank Withdrawal from a Subcritical or low Power l

The core power is assumed at a critical zero power startup condition.

Peak RCS pcessure analysis is performed assuming the RCPs are operational to minimize thermal feedback during the power excursion. Reactor trip is assumed on high power range flux trip.

DNB analysis will be performed using the SCD method except when the potential for bottom-peaked power distributions exists. In such event, since SCD is not applicable, DNBR analysis is performed using the W-3S CHF correlation in the traditional manner accounting for uncertainties explicitly. Thus the input selection criteria described below is only applicable when the SCD method is used.

In order to determine the statepoints to be used in the DNB analysis, the initial conditions- for the SCD treated parameters for the cases are ret at nominal conditions for this power with three RCPs in operation. To minimize the PZR pressure increase, low initial PZR pressure and level is assumed.

Three RCPs, a minimum number required for the modes of operation applicable for this transient, are assumed operational to yield low flow. Reactor trip is assumed on high power range flux trip. -

3.3,4.2 Uncontrolled Bank Withdrawal from Power For peak RCS pressure analysis, in order to avoid trip on high flux, the transient is initiated from low power. The SG level is assumed high and a high amount of SG tube plugging is assumed in order to minimize primary-to-secondary heat transfer.

In order to determine statepoirts to be used in DNB analysis using the SCD method, the initial conditions for the SCD treated parameten s for the cases are set at the nominal conditions corresponding to each of the power levels, which span the full spectrum, for which this event is analyzed. The steam 15

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~

generator level is assumed to be high in an effort to minimize the primary- (

to-secondary heat transfer. Analysis is performed with and without PZR l sprays and PORVs.

3.3.4.3 Control Red Misoceration Transient systems analysis is not performed for the statically misaligned control rod event. Steady-state three-dimensional power peaking analyses are performed to assure that the resulting asymmetric power distribut h will not result in DNB.

3.3.4.4 Sinale Rod Withdrawal DNB analysis will be performed using the SCD method.

The SG mixture level is assumed high to maximize the secondary pressure and minimize the primary-to-secondary heat transfer. High SG tube plugging is

-assumed to minimize the primary-to secondary heat transfer.

Reactor trip is assumed on one of four furictions; 010T, OPDT, PZR high pressure and powcr range high flux. In order to delay reactor trip on high PZR pressure, the PZR heater: is assumed to be in manual. Similarly the PORVs are ass"~ad disabled in order to delay reactor trip on OTDT and high PZR pressure.

eedwater control is in automatic to prevent SG level trip.

AFW is assumto disabled. Turbine trip is assumed on reactor trip.

3.3.4.S Startuo of an___Jnactive RCP at an incorrect Temnerature DNB analysis will be performed using the traditional method. A two-loop model will be used because of the loop asymmet,y.

The initial indicated power level is set to delay or prevent reactor trip from a low flow trip .setpoint. The core bypass flow is assumed to be high to minimize the core flow to maximize the heatup. Similarly the RCS flow in the )

three unaffected loop is '. a minimum flow allowed by Technical Specification.

l i

16

The three unaffected RCPs are modeled assuming a constant speed through the transient. The RCP that is initially inactive is modeled with a conservative

( speed vs. time controller.

The SG level control is assumed to be in automatic mode to minimize the probability of trip on low-low SG 1evel. Turbine trip is assumed to be in manual.

The input selection and transient assumptions as described in the topical report for this category of events is acceptable.

3.3.5 increased in Reactor Coolant Inventory Inadvertent operat;on of ECCS during power operation is the only transient analyzed. The DNB results of this transient are bounded by the inadvertent opening of a PZR safety or relief valve transient.

Notwithstanding the qualitative argument provided by DPC for not analyzing this event, DPC nevertheless presented the analytical methodology used for this analysis, should reanalysis become necessary in the future.

DNB analysis will te performed using the SCD method. -

A maximum safety injection flowrate with a conservatively high boron concentration is assumed to yield the most limiting transient response because it minimizes power and thereby maximizes the amount of ECCS which car.

be injected. In order to minimize the delay in the delivery of the barated water, no credit is assumed for the purge volume of unborated water in the injection line.

Reactor trip is assumed on low PZR pressure after an appropriate delay time.

The steam line PORVs and condenser steam dump are assumed to be unavailable to maximize secondary side pressurization and minimize the primary-to-secondary heat transfer, also tending to maximize primary fluid volume.

17

. . . . . _ _ _ - _-_____--_______ _ _ _ _ _ _ - - . - _ - - - - - - . . - - - - . O

Turbine trip _is assumed on reactor trip.

The input selection and transient assumptions as described in the topical report for this category of events is acceptable.

3.3.6 Qecrease in Reactor Coolant Inventory Inadvertant opening of a pressurizer safety or relief valve and steam generator tube rupture events are the two transients analyzed in this category. The steam generator tube rupture event is beyond the scope-of this review. Therefore, the inadvertent opening of a PZR safety or relief valve was reviewed.  !

In order to determine statepoints to be used in DNB' analysis using the SCD method, the pressurizer liquid level is assumed to be high to maximize the primary pressure decrease, which maximizes the added coolant inventory.

Reactor trip is credited. The turbine trip is assumed on reactor trip without delay to minimize post-trip primary-to-secondary heat removal.

The input selection and transient assumptions as described in the topical report for this category of events is acceptable.

4.0 CONCLUSION

S DPC topical report DPC-NE-3002 and its supporting documents, including the DPC response.s to questions, were reviewed.

Review of the subject topical report focused upon evaluation of acceptanility of the RETRAN models for the type of analysis generally described on the subject topical report. The topical report was further reviewed to assure that the application of the DPC's DNB methodology was consistent with the centents of DPC-NE-2004 and DPC-NE-3000 and acceptable. The review, therefore, included identification - of the SCD transients and evaluation of DPC's selected initial and boundary conditions in the systems analysis which was used to determined the statepoints for the DNB analysis.

18

As stated earlier, steam line break, rod ejection, dropped rod, steam generator tube rupture and boron dilution events were not part of this review (see also Section '.1).

Subject to the foregoing, DPC's approach to FSAR Chapter 15 transient analysis, as documented in DPC-NE-3002 and its supporting documents, was generally found to be acceptable subject to the following conditions:

1. DPC's Statistical Core Design methodology treat seven state variables as key parameters. Four of these variables were accounted for in this topical report. Of the remaining parameters, the power factors are also input items for systems analysis, which was not presented in the topical report. Similarly, reactivity feedback was not discussed in this report. Both of these parameters can significantly influence the course of the transient. Therefore, when application of the philosop'.iical spproach reported in this topical report is made and submitted for NRC review and approval, review should be made of the modeling of power and reactivity feedback, and to assure that such modeling has no adverse impact on the other modeling described herein.

2. Validity of DPC's assumption of 120% of design pressure as part of the acceptance criteria for Reactor Coolant Pump t.ocked Rotor should be _

determined by the NRC staff.

3. No justification was presented for trip and actuation times assumed in the Feedwater System Pipe Break event analysis. Such justifications must be presented when this methodology is applied.

4. OPC documented interit to perform parametric studies in order to select conservative scenarios or assumptions throughout the subject topical report. Therefore, such parametric studies must be presented when this methodology is applied.

19

_ ._ ______________ _. _ A

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5.0 REFERENCES

(Amended to reflect transmittal dates) =

1. Letter, H. S. Tuckman (DPC) to NRC, "FSAR Chapter 15 System Transient Analysis Methodology," DPC-NE-3002, August 30, 1991.

2. Letter, C. O. Thomas (NRC) to T. W. Schnat: (UGRA),Septembera,1984, (Transmittal of RETRAN-02 Safety Evaluation Report).

3. Letter, C. E. Rossi (NRC) to J. A. Blaisdell (UGPA), May 1, 19' (Transmittal of VIPRE-01 Safety Evaluation Report).

4 Letter, H. B. Tucker (DPC) to NRC, " Thermal-Hydraulic Transient Analysis Methodology," DPC-NE-3000, September 29, 1987.

5. Letter, H. B. Tucker (DPC) to NRC, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01,"

DPC-NE-2004, January 9, 1989.

6. -Letter, M. S. Tuckman (DPC) to NRC, " Supplemental Information to Assist in Review of Topical Reports DPC-NE-3000 and DPC-NE-2004," August 29, 1991.

7. Letter, H. B. Tucker (DPC) to NRC, " Handouts Presented in the October 7 & 8, 1991 Heeting with NRC Staff and Contract Reviewers," October 16, 1991.

8. Letter, H. B. Tucker (DPC) to NRC, " Final Response to Questions Peoarding the Topical Reports Associated with the MIC8 Reload Package," November 5, 1991.

9. Letter, T. A. Reed (NRC) to H. B. Tucker (DPC), Safety Evaluation on lopical Report DPC-NE-2004, " Core Thermal-Hydraulic Methodology using VIPEE-01," November 15, 1991.

10. Letter, T. A. Reed (NRC) to DPC, Safety Evaluation on Topical Report DPC-NE-3000, " Thermal-Hydraulic Transient Analysis Methodology," November 15, 1991.

11. Letter, T. A. Reed (NRC) to H. B. Tucker (DFC), Safety Evaluation on Topical Peport DPC-NE-3001, " Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology," November 15, 1991.

20 1

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DUKE POWER COMPANY

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McGUIRE NUCLEAR STATION CATAWBA NUCLEAR STATION I'

FSAR CHAPTER 15 SYSTEM TRANSIENT A!JALYSIS METHODOLOGY-l ;.

! DPC-NE-3002-A i

NOVEMBER 1991 1

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. Nuclear Engineering' Group Nuclear Services Division Nuclear Generation Department

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[7 This report documents the conservative modeling assumptions used by Duke I: power Company in performing the t;sSS primary and secondary system l: analyses of FSAR Chapter 15 accidents. It covers all applicable non-l- LOCA' accidents in Sections.15.1 through 15.6 of the PSAR except those already addressed in Duke Power company topical report D PC-t!E - 3 0 01. The l ,- areas discucsed are nodalizatior,, initial conditions, boundary je conditjons, Inodeling of the process instrumentation and control systems, I the Reactor Prctection System, the Engineered Safety Features Actuation l System, and availability of.other important systems (.nd components, i-l l

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FSAR CHAPTER 15 SYSTEM TRANSIENT A!IALYSIS METHODOLOGY Table of contents

1.0 INTRODUCTION

2.0 INCREASE IN HEAT RbN".u BY THE SECONDARY SYSTEM 2.1 Feedwat er Cyrtem Malf unct ions That Result In A Pedurtion In Feedwater Temoerature 2.2 Feedwater 9vstem Maltuno#_ ion causina an Increase in Feedwater Flow 2.2.1 Nodalization 2.2.2 Initial Conditions -

2.2.3 Boundary Conditions 2.2.4 Control, Protection, and Safeguards Systems Modeling 2.3 Excessive Increase in recondarv 9 team Flow 2.3.1 Modal 1:ation 2.3.2 Initial Conditions 2.3.3 Boundary Conditions 2.3.4 Control, Protection, and Safeguards Systems Modeling 2.4 Inadvert ent Ocenina of a 9t eam Generator Pelief or Safety Valve 3.0 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 3.1 Turbine Trin 3.1.1 Peak RCS Pressure Analysis 3.1.1.1 Nodalization 3.1.1.2 Initial Conditions 3.1.1.3 Boundary Conditions 3.1.1.4 Control, Protection, and Safeguards System Modeling 3.1.2 Peak Main Steam System Pressure Analysis 3.1.2.1 Modalization 3.1.2.2 Initial Conditions 3.1.2.3 Boundary conditions 3.1.2.4 Control, Protection, and Safeguards System Modeling 3.2 Lors of Mon-Emernenov Ac Power To The Fration 3.2.1 Pe k RCS Pressure Analysis 3.2.1.1 Nodalization 3.2.1.2 Initial Conditions 3.2.1.3 Boundary Conditions 3.2.1.4 Control, Prctection, and safeguards System Modeling 3.2.2 Peak Main Steam System Pressure Analysis 3.2.2.1 Modalization 3.2.2.2 Initial Conditions 3.2.2.3 Boundary Conditions 3.2.2.4 Control, Protection, and Safeguards System Mode'ing 3.2.3 Core Cooling Capability Analysis 3.2.3.1 Nodalization 3.2.3.2 Initial Conditions 3.2.3.3 Boundary Conditions ii

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3.2.3.4- Control, Protection, and Safeguards System Modeling 3.3 Loss Of~Hormal Feedwater-3.3.1 Peak RCS Pressure Analysis 3.3.1.1 Nodalization 3.3.1,2 Initial ~ Conditions 3.3.1.3- Boundary Conditions 3.3.1.4 Control, Protection, and Safeguards System Modeling  ;

3.3.2 Peak Main Steam System Pressure Analysis 3.3.2.1 Nodalization 3.3.2.2 Initial conditions 3.3.2.3 Boundary conditions

.3.3.2.4 Control, Protection, and Safeguards System Modeling 3.3.3 Core Cooling Capability Analysis 3 . 3 .-~ 2 .1 Modalization 3.3.3 2 . ' Initial' Conditions 3.3.3.3 Boundary Conditions 3.3.3.4 . Control,' Protection, and Safeguards System Modeling.

.3.4 .Epodwater ristem Pine Break 3.4.1 Short Term core Cooling Capability 3.4.1.1 Nodal 12ation 3.4.1.2 -. Initial Conditions

-3.4.1.3 Boundary Conditions 3.4..,4 control, Protection, and Safeguards System Mode 3.4.2 Long Term' Core Cooling Capability 3.4.2.1 Nodalization 3,4.2.2 Initial. Conditions 3.4.2.3 Boundary. Conditions

. 3.4.2.4 Control, Protection, and Safeguards System Modeling 4.0 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE

.1 . Partial Loss of Forced Reactor coolant Flow 4.1,1- -Noda11zation-

.4.1.2 Initial Conditions

'4.1.3 Boundary Conditions 4.1.4 Control, Protectioni-and Safeguards System Modeling

.4.2 Comnlete Loss Of Forced Reactor-Coolant Flow 4.2.1 'Nodalization

-. 4 . 2 . 2 Initial Conditions-r4.2.3- Boundary Conditions 4.214L -Control,- Protection, and Safeguards System Modeling

- 4.3 Resctor Coolant Puro Locked Rotor

.4i3.1 Peak RCS Pressure Analysis 4.3.1.1 Nodalization

. 11.3.1.2~ Initial Conditions

- 4 ~. 3 .1. 3 - Boundary Conditions 4.3.1.4- Control, Protection, and Safeguards System Modeling 4.3.2 _ Core Cooling capability Analysis 4.3.2.1 Nodalization 4.3.2.2' Initial Conditions

-4.-3.2.3 Boundary Conditions 4.3.2.4 Control, .Ptotection, and Safeguards System Modeling 4.3.2.5 'Other Assumptions iii

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5.0 REACTIVITY AND PO' DER DISTRIEUTION ANOMALIES 5.1 Uncontrolled Bank Withdrawal From a Fubciltical or Low Power 5.1.1 Peak RCS Pressure Analysis 5.1.1.1 Nodalization 5.1.1.2 Initial Conditions 5.1.1.3 boundary Conditions 5.1.1.4 Control, Protection, and Safeguards System Modeling 5.1.2 Core Cooling Capability Analysis 5.1.2.1 Nodalization 5.1.2.2 Initial Conditions 5.1.2.3 Boundary Conditions 5.1.2.4 Control, Protection, and Safeguards System Modeling 5.1.2.5 other Assumptions ~

5.2 Uncontrolled Bank Withdrawal at Power 5.2.1 Peak RCS Pressure Analysis 5.2.1.1 Nodalization 5.2.1.2 Initial Conditions 5.2.1.3 Boundary Co? ditions 5.2.1.4 Control, Protection, and Safeguards system Modeling 5.2.2 Core Cooling Capability Analysis 5.2.2.1 Nodalization 4 5.2.2.2 Initial Conditions 5.2,2.3 Boundary Conditions 5.2.2.4 Control, Protection, and Safeguards System Modeling 5.3 Control Pod Miscoeration 'Statica11v Misalianed Rod) 5.4 Control Pod Misoneration (Finale Rod withdrawal) 5.4.1 Nodalization 5.4.2 Initial conditions 5.4.3 Soundary Conditions 5.4.4 Control, Protection, and Safeguards System Modeling -

5.5 Startun Of An Inactive Poactor Coolant Purn At An Incorrect Temnerature 5.5.1 Nodalization 6.5.2 Initial Conditions 5.5.3 Boundary Cond'tions 5.5.4 Control, Pre action, and Safeguards Systems Modeling 5.6 CLE;E. Malfunction That Results In A Oecrease In Boron Concentration In The Peactor Coolant 5.6.1 Initial Conditions 5.6.2 Boundary Conditions 5.6.3 Control, Protection, and Safeguards System Modeling 5.7 Inndvertent Loadina and Operation of A Fuel Assemb1v In An Imoroner Position 6.0 INCREASE IN REACTOR COOLANT INVE!CORY 6.1 Inadvertent Operation of ECCF Ourina Power Operation 6,1.1 Nodalization 6.1.2 Initial Conditions 6.1.3 Boundary Conditions 6.1.4 Control, Protection, and Safeguards System Modeling i iv

- _ . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -------------A

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l 7.0 - DECREASES IN REACTOR COOLA!!T IlWENTORY 7,1- intivertent Orenina of a Pressurirer safety or Relief 7.1.1 IJodalization' 7.1.2- -Initial-Conditions 7 .1. 3 . Boundary Conditions 7 1.4 Control, Protection, and Safeguards Systems Modeling

-7.2 steam cenerator Tube Pupture 7.2.1 Core Cooling Capability Analysis 7.2.1.1 tJodalization 7.2.1.2 .' Initial-Conditions

'7.2.1.3 Boundary Conditions 7.2.1.4 Control,. Protection, and Sa'eguards System Modeling 7 . 2 . 2 -- Offsite Lose Calculation Input Analysis 7.2.2.1 tiodalization' 7.2.2.2 Initial Conditions 7.2.2.3 Boundary Conditions j 7.2.2.4 Control,' Protection,.and Safeguards-System Modeling 8.0

9.0 REFERENCES

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1.0 IHTRCDUSTICM This report documents the conservative modeling assumptions used by Duke Power Company in performing the NSSS primary and secondary system analyses of FSAR Chapter 15 accidents. It covers all applicable non-LOCA accident s in Sections 15.1 through 15.6 of the FSAR except those already ad. dressed in Duke Power Company topical report DPC-NE-3001 (Reference 1), which are steam system piping failure (FSAR Section 15.1.5), control rod misoperation (dropped rod, rod group, or rod bank, l FSAR Sections 15.4.3atb), and rod ejection (FSAR Section 15.4.8) 1 The only accidents categorized as not applicable are those which 1) do not apply to McGuire and Catawba (FSAR Sections 15.2.1, 15.5.3, 15.5.4, and 15.6.6), 2) involve no system thermal-hydraulic analysis (FSAR Section 15.6.2), or 3) the current McGuire and Catawba licensing bases regard as _

l being bounded by another accident (FSAR Sections 15.2.2, 15.2.4, 15.2.5, 15.3.4, and 15.5.2). The assumptions discussed in this report are specific choices about the use of the models described in general in DPC-NE-3000 (Reference 2). The areas discussed are nadalization, initial conditions, boundary conditions, modeling of the process instrumentation and control systems, the Reactor Protection System, the Engineered Safety Features Actuation System, and availability of other important systems and components.

The discussion of the nodalization employed in analyzing a particular accident focuses on two main areas. First, the symmetry of the accident is examined to determine whether it affects all Reactor Coolant System (RCS) loops in approximately the same manner , justifying the use of a single RCS loop model, or whether one or more loops must be modeled separately to conservatively model differential effects of the accident on them. Second, the level of detail of the models described in Ref er ence 2 is examined to determine whether they are appropriate f or each analysis. In most cases the modeling described in Reference 2 is appropriate. Any inadequate modeling would be upgraded on an accident -

specific basis to ensure conservative modeling of the physical phenomena requiring a more detailed model. Modeling regarded as excessively detailed, considering the importance of that area of the system in the particular accident, might be simplified to reduce the computational costs or the effort required to simulate that section of the model.

The analyses covered by this report are intended to be valid, unless stated otherwise, for both the McGuire and Catawba Muclear Stations.

For each analysis, the differences between the two stations and between the two units at a given station, as discussed in Section 3.1.C of Reference 2, are considered. A bounding " unit" is selected considering how these differences affect the margin to each acceptance criterion of the accident being analyzed. In some cases this is an actual unit, e.g_, the use of Catawba Unit 2 because its steam generator inventory as a function of power is different from the other three units. In others it is a superpesition of limiting characteristics from more than one unit, e.g., using steam line safety valve banks which correspond to the two lowest setpoint McGuire valves and tne three highest setpoint Catawba valves since this artificial bank has a smaller relief capacity than the actual banks at either station. In the future such combined 1-1

_- __ _ _ _ - _ _ _ _ _ _ _ _ - _ - - _ - _ _ - -_ _____ -_ - _ _ - _ _ _ _ _ a

analyses might be redone separately on a more plant specific basis to gain margin.

The values for relevant plant parameters at the start cf each accident are determined through the following process. First, ae value for a given parameter is determined considering normal and off-normal plant operation, Technical Specification limits, and mode of parameter control (whether controlled by an automatic system or manually by the operator).

Since many of the important. parameters are functions of reactor power, some of the parameter value choices are made to be consistent with the initial power level for the accident. Once the parameter value is determined, a method is used to account for uncertainties in this value due to controller tolerance (either manual or automatic) or instrument uncertainty. This method might be an explicit adjustment to the initial l value'itself or an accounting for the unce-'ainty in other affected l parameters, such as DNBR limits or reacto crip setpoints. Parameters for which an uncertainty adjustment is made are listed in Table 8-1.

1 The boundary conditions which affect the course of the transient are modeled to ensure a conservative result. Boundary conditions include:

1) Flows to and from plant components not explicitly modeled, e.g.,

Emergency Core Cooling System (ECCS) flow rate ar a function of ECCS configuration, RCS back pressure, ECCS suction source l temperature, pressure, and boron concentration, pump motor starting time, and any postulated pump degradation i i

i

2) Releases through pipe breaks and open valves, including the I effects of critical flow

3) Timing of manual actions

4) Timing of automatic actions, including the effee s of setpoints, setpoint tolerances, and-the uncertainties in monitored parameter signals The modeling of boundary conditions is very accident specific and is discussed in detail under each accident.

The plant control systems modeled for accident analyses are described in Sections 3.1.4 and 3.2.4 of Reference 2. Only those control systems which have an important effe-t on the course of the accident are considered. If the operation of a given control system would mahg the accident more severe, that system is assumed to function normally. If its operation would make the accident less severe, the system is not  !

assumed to function. The Reactor Protection System (RPS) and the Engineered _ Safety Features (ESP) are described in Sections 3.1.5 and I 3.2.4 of 9 2. Only those safety systems which have an important effect on the course of the accident are considered. The most limiting single active failure of a component to perform its safety functions is considered in accordance with Appendix A to 10 CFR 50. l In general, no credit is taken for components which are not safety grade, although a penalty for their operation might be taken as i

l 1-2 l

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der.cl ll>e<l above S it:1latly, the availabi11.f of ncn-saf at y syst er: 5 an;l ccalonents, e.g., l eact ot coolant puty c (ECis), pressurtret he rs t e r s ,

L non-eniorgency AC power, and insttu wnt air, is only assur:,ed if such availability would ruke the accident votse.

The lirt c f <tr.oun pt ions for the accident s is suwutized in Table 8-1.

L:ach accident durct lpt ion gives t he t elevant subset cf t ho s e a s s u::ipt i otic applicat-le for a 1 at t icular accident and diccusses their Lases.

9 1-3

.._ .. J

.; . 0 Itt i ElwE IN Hi:AT l-ItoVid. W THE PL CNC;s1T YETIM L

2.I lj d otEi :);t C lbliun=t.13n..That 19;;jlt in A F.e N"ti n In k' led L Y T' ' Qf ' l l . t A Fe edaat.et Lyst en no!!unct ion that Imult s in a der t e nz e in foet/attI t m}+1atute wilI cwe n 1r ct eaw in cot e Imt by de c t eE 17.J t e a c t. c t

oolont ternperatute l h,m i c a l ly, as the cooler f oodwat er r educt s the 1 < v t et coolant t o r+e t at u r e , positive t eact ivit y will be incelt6 d due to the effect at a negat ive A>%2ator t en pri at u r e coe! ! i c i e r.t . im ulat ing that the Fod ContIol Syst 4 m in in autor,atic control, control iOA would be WithdtaWn ab Rc5 IOdyttatutb dect 6 dUed, iru et t ing addit.ional [405 it ive 2 eact tvit y. The net <eftect on the Ecr due to a 1 educt ion in Ieedwat er t en per at ur e would 1+ J irni l a r t o t he ef f ect of increasing f eedwat et flow _

cr inc reasing seccrrJa ry c t esn. Ilowt the reactor will teach a now equillt t iurn corrli t ion at a power level cor responding t o t he new sicam gener at or AT.

A Feedwater System malf unct ion that resultc in a decr eam .a f e+- dwat t i tempetatute can 1e init lat ed f I on the fol3owing typer of eventt -

spur inus act uat ion of a f eedwat et heat er bypans valve, int ert upt ion of st eam ext laction f low t o a f(edwater heater (s), spurious startul, of a single auxiliary f eedwat et punp, failure of a single f eedwat er heat et dr ain purnp or failure of all f eedwat et heatet drain pumps. The above events ato examintd, with t he ruost limit ing deter mined to be a opurious act uat ion of a f eedwat et heat er bypast valve However, undet the cut t ent Duke Power Company inet hod of analynis, thia accident it bounded by quant itat ive analysic of the increase in icedwat er f low event or t h.

excescive jncrease in secondat/ steam flow event. These evente bound the reduction in feedwater tempetature event by producing a greatet RC" cooldown. The applicable acceptance criterion is that fuel cladding int egr ity shall be maintained by ensuring that t he ininimum DNbh temains above the 9 '3 / 9 5 I ' .14 R l i m i t based on acceptable correlations.

J.2 Feetcat er cy r t ers Malfuncticn Onurina an Increare in Dee lnat c F] my Thu malfunctions considered are 1) the full opening of a single main feedwater conttel valve, 2) an inct3ase in the t, peed of a single main feedwater pump, 3) the spurious startup of a single au).iliary feedwater pump, or 4) a rulIunction which af f ect s more than one loop. The latter scendt io has been ident if ied only recent ly and is curr ent ly being evaluat ed t o determine applicability to McGuit o an.1 Catawba. The limit ing scenarlo t rot., among those listed above is evaluat ed to demonstrate that fuel cladding integrity is maintained by ensuring that the minimum DNi1R remains above t he 95/95 DNER limit based on acceptable cor relations using t he St at istical Cole Cesign Methodology 2-1

- _ _ _ _ _ . _ _ _ _ _ - - - _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________J

. )

2.2.1 Hodalization Of the event s identified in the previous sect ion, the latter, the multi-loop malfunction, is expected to be the most limiting, and is therefore the one that is discussed. This transient affects all loops eaually and would t het ef or e be analyzed wit " single-loop NSCS system model (P.oference 2. Section 3.2). If niost litciting transient is not determined to be one which af f ees all loops equally, a tuultiple loop toodel would be selected appropriat ely. The pressurizet modeling includes the uno of the local conditions heat transfer option for the vessel conductors.

2.2.2 Initial Conditions ggr e Pown Lovel liigh initial power level maximizes the primar/ system heat flux. The uncertainty in this parameter is accounted for in the Statistical core ,

DcGign Methodology.

M er curiret Piernure  ;

The notoinal pressure cor responding to f ull power operation is assumed, ,

with the pressure initial cotittion uncertainty accounted for in the Statistical Core Desagn Methodology. ,

M er c u rir,t r Level Since this accident involves a reduction in RCS volume due to coolant contraction, a positive level uncertainty is applied to the nominal programned level to minimize the initial- prewarizer steam bubble volume and therefore ma.ximize the pressure decrease due to contraction.

• React or veerel Averade Temoerature The nominal temperature corresponding to full power operation is assumed, with the temperature initial condition uncertainty accounted for in the Statistical Core Design Methodology. r Rcs Fluy The Technical Sper"fication minimum measured flow for power operation is assume (. since low W ow is conservative for DNBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design iethodology.

Core Pyr u s F) cx The nomical calculated flow is assumed, with the flow uncertainty accounted for.in~the Statistical Core Design Methodology.

i Pteam Generator Level A negative level uncertainty is assumed to maximize the margin to a high-high steam generator narrow range level reactor trip due to any

-temporary steam /feedwater flow mismatch. This maximizes the duration of the overcooling before it is ended'by foodwater isolation.

3 2-2

3 1 To... g ure A low irrit ial t e:tpet ut ur e is as sumed t o inaximite the gap ecnduct ivity calculated for st eady-st at e condit tens and used fet the subsequent transient. A high gap conductivity rninimizes he fuel heatup and at t errl ant negative t eactivit y insertien caused by t he I.ower incr ease.

This baker the powel increase more rever e and i r. theitsfore cor1Gervative Ior !!mR evaluatlen.

rt er cernger Tube PluggiIc In older to maximize the ef f ect s of the incr eased secondary syst ern heat removal, no t ube plugging in assumed.

2.2.3 boundary Conditions thin reetat er Ple-

• A conservatively lat ge step change in rnain f eedwater flow to all cteam generatorc is assumed at the initiation of the transient.

2.?.4 Control, protection, and Safeguards Systems IP>deling ynu;t ei irio The pet t inent teact or trip tunctions ar e t he low-low cten generator level, high Ilux and overpower AT. The safety analysis cetpoint of the init ial condition f or the Inonitored parameter contains an allowance for inoasur ement instrumentation uncertainty and sotpoint setting tolerance.

11erruti~er Level control flo credit is taken 101 pressurizer level centrol system operation to compensate for the d m ccuritation which accompanies RCS volume ch t'i nkag e .

Red control This accident will resuit in a decrease in RCS t emperature. With the Rod Control System in manual control, the reduced temperature will cause a posit.ive t eactivity insertion through the negative moderator temperaturo coefficient. With the Rod Control System in automatic conttol, in which the reactor vessel average temperature is maintained at a progremmed value, the control rods will cause a positive reactivity insettion as they are withdrawn in an attempt to maintain this t einper atur e . Both cases are analyzed in order to ensure that the worse one is determined.

Turbine contrtl The tutbine is modeled in the load control modo, which is described in Section 3.2.5.1 of Reference 2. In this rhode any decrease in steam pressure, due for example to a shift from latent to censible heat transfer because of the overfeed, would be compensated for 1.y an opening of the turbine control valves to naintain impulse shamber precsure at the programmed value.

2-3

.. __.J

[.ux i 11 n v F _. o 1 :4 1.1 AFW flow it cle12ted, aftet t he a[ } o} t t at e Te chnical r}ecificatioh l et parae t into delay, when t he ratet y analysis vslue of the l ow - l ow s t e sm genetatat natIow Iange leve1 set point is aeached. In os det- ta rinsuare tbe pcst-t11p cteam generator heat 2ere'al, the minimum aux!1ialy

!endwatet ilow is a s o u:w d .

LL1-1 n.

Tiip Tutb1ne t11p ic etedited, at t er the op, I c pt iat e Technical Epeci!Icatton 1 esponse t true delay, on high-high ct eJen ge net at or nat iow I Ange level.

lid ut t i h elaticn Feedsatet it,olation is credited, after the ai rt opr iat e T*:c hni c a l rpocification tespcnse t une del ay , cn high-high ctearn gener at o! hattow range level.

2,3 &..e3".- Inciease in En" A " > "' en F1 m The accident analyzed is a step increase in secondary steam flow of a .

tuagnit ude equal t o t hat f or which the React et Cent rol Syst ent is designed, 10% of licensed cor e t herrnal powel . Incr eases of larger magnitude at e discusced in Section 2.4 and in ch'ipter 5 of Fef erence 1.

The accident in ana.itn 1 to demonst rat e t hat fuel cladding integrity is tuaintained by ensurird that. the Ininimum D!mR t emains above the 95/95 D:IbR lin-it based en acceptable correlations. The na ninum D!mR is det ettnined using the Statistical Core Design Methodology.

2.3.1 IJodalization The accidtnt analyzed is an excescive inctease in t.econdary steam flow at po*er. Flow incteases f rom a zero power initial condit ion are evaluated in Section 2.4 and in Chap'er 5 of Reference 1. Per Feference 3, Section 15.1.4, t he power lovel analyzed f or this accident. should be 102% of licensed core thermal power for the number of loops initially assumed to be opetating. At power, the Technical Specifications requit e all f our loops t o be operating. Therefore full power is assumed as the initial condition. An increase in staam flow to t he tut bine would p affect all loops equally, therefore, a single-loop !!SSS system model (Reference 2, Section 3.2) is used. Tne ytessurizer modeling includes t he use of the local condit ions heat transfer option ior the vessel conductor 2.3.2 Initial Conditions core Pown_ Level Per Reference 3, Sect ion 15.1. 4, the power level analyzed for this x accident should be 102% of licensed core thermal power for the number of ,

loops initially assured to be c.perating. At power, the Technical s Specifications require all four loops to be operating. Therefore full p power is assumed as the initial condition. The uncertainty in initial ,w-power level is accounted f or in the St atistical Core Design Methodology.

2-4 u __ _- - ._ _

) 11 e n u r 12 ei Pierruie The nominal ptei.sute corterpond1Dg to full power operation is asrumed, with the p2 essure initial condition urcettainty account ed f or in the Ctatictical Core Design liethodology.

h ulizei Level since thin accident involves, particularly for the manual Rod control System operation scenario, a reduction in RCS vclume due t o coolant contraction, a positivo level uncertainty is assumed to minimize the initial pressurizer steam bubble volume and therefore maximize the prescure decrease due to contraction.

Egg t er Venol Averace Teni eratute The nominal temperaturo corresponding to f ull power operat ion is assumed, with the tempetature initial condition uncertainty accounted for in the Statistical Core Design Methodology.

PCE Flow The Technical crecification minimum measured flow for power operation is assunied since 'ow flow is conservative for DIER ovaluation. The flow initial condition uncertainty in accounted for in the Statistical Core Design Methodology.

Corn tvim e Flew The nominal calculated flow is assumed, with the flow uncertainty accounted tot in the Statistical Core Design Methodology.

EteAm Gencia;n_, Level The results of this transient are not sensitive to the direction of steam generator level unco;tainty ao long as the trans16nt level response is kept within the range that avoids protection or safeguards actuation.

Fue 1 Tomt'ei at ur e A low initial temoeratute is assumed to maximize the gap conductivity calculated for steady-state conditions and used for the subsequent transient. A high gap conductivity minimizes the fuel heatup and attendant negative reactivity insertion caused by the power increase.

This makes the power increase more severe and is therefore conservative for DNB evaluation, rtenn Generator Tube Plugaina In order to maximite the ettects of the increased secondary system heat removal, no tube plugging is assumed.

2.3.3 Boundary Conditions Main Pt erm Flow A step change in main steam flow to the turbine equal t o 10% of full power flow is assumed at the initiation of the transient.

2-5 a

2 . '! . 4 Cont t ol, Pr ot ect ion, and I;afegualds Fyst ems Podeling

&_-r t e t Trin The reactor is not expect ed t o t rip ior this transient. H > wever ,

t eact or t r ip ir credit ed, after the appropriate Technical Specif icat icn tespcnse time delay, if the safety analycis setroint is exceeded for any react or trip function.

6 trurinr Level ce ntrol 11 0 credit is t aken f or pressurizer level control system cretation to conpen sat e for the deprecsurization which acccep nies ECS volume chtinkage.

Ste_im Line PORVs and ccndenser St er Den While the steam line PORVs and steam dump might be a soulce of the incteased steam flow in this postulated accident, the case analyzed assumes the increased flow exits to the turbine.

Steam Cenerater Level centrol The results of this transient are not sensitive to the mode of steam generator level control as long the level ts kept within the range that avoids protection or saf eguards actuat lon.

UW Pure roeed cent r ol The r esult s of thin ttansient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.

Rod control This accident will result in a decrease in RCS temp,erature. With the Rod Control System in manual control, the reduced t.emperature will cause a positive reactivity insertion through the negative moderator temperature coefficient. With the Rod Control Systcm in eutomatic control, in which the reactor vessel average temperature is maintained at a programmed value, the control rods will cause a posit.1ve reactivity n insertion as they are withdrawn in an attempt to maintain this temperature. Both cases are analyzed in order to ensure that the worse one is considered.

Tutbine centrol The turbine is modeled as described in Section 3.2.5.1 of Feference 2, with a step increase in flow rate at the beginning of the accident. ,

Any11iary Feedwater AFW flow would be credit ed, after the appropriate Technical Specification response time delay, when the safety analysis value of the low-low steam generator level setpoint is reached. However, the k paramoter of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has olapsed. Therefore, no AFW is actually delivered to the steam generators.

2-6

.4 In.dvei t t nt ni;,1 nf a it* ne rw ! ': 1 Eelief --- ? a ftt' Valvt s

This accident it s i:ci l a r i n nam t lespects t o t he s t > urn line L: e ak

• accident analyred in Charter 0 of F ef er rnre 1. If the 5 n adver t ent ly co ned valve will not receat, and cannot t ., ioolat ed by c1r ning a valve in aurier with it, the eI!ect as the sanc as a1 Ile 1 leak in t he ese location aid wit h t he ser effective flow area. becaure the ste m line r at et y valvec and t he steam line power -o; ot ated r elief valve. ( lak'/ s )

ate locat ed u;'c t t oara of the Malve, a uten line isolation actuation, wit h or without a failute of a single IC I V , would tesult in t he continued blowdown of the ot ten. genez at or with t he f ailed valve.

Lecause of the similarities between thic accident and t he steam line br eak accident. , the inadvertent evening of a stem generator relief or saf et y valve is not quant it atively analyzed. Undel the current Duke -

l'ower Company n.ethod of analysis, this accident is boundod l'y the quantitative analysis of the steam line bleak accident. The atplicable acceptance etitetion is that fuel cladding integrity shalI be maintained t y ennut ing t hat the mininoim D!ER remains above t he 9 5 / 9 5 Dt.'bh limit based on acceptable correlation 7 This criterion is satisfied by con,parison t o the D!!UR t esult s for the more limiting st eam line tr eak transient. In Section 5.4.2.2 of Refetonco 1, it is shown that HII4 does not occur for the steam line break accident. Cince DIW is not unacceptable f or t he Uteam line break event, it is possible that future steam line break analyses might conclude that NIB occurs f or that event .

If this occuin, the analytical methodology for the steam line bleak analysis will 1,e applied t o an analysis of the inadvertent opening of a steam generator relief or cafety valve, wit h an appropriate adjuctnent t o t.he btvak fJow atea.

2-7

- _ _ _ _ _ _ -_ - - D

3.0 UECRF.ASE 111 MEAT R12DVAL DY TlIE EECCl DARY SYSTE!!

.$ .1 Pn t i r Ti ip The t ut t>i ne trip event causes a loss ot leat sink t o t he pr in ar y sys'.em.

The Inisnatch between powel g ner at ion in t he pr imar y cyst s m and heat iemoval by t % secondary system caurec t e:4 + r a t u t e and ptessure to increase in t he pt imary and se:ondary unt il reactor trip and/or lift of the prescutizer safety valves and nain steam saf ety valvec. The t r ans i ent is analyzed t o ensure that bot h the peak Reactor Coolant cyntom pressute and t he l eak Main St ema System pressute t emain below t he acceptance cr it er ion of 1101 of design pressure. Peak RCS pressure and peak M stn St eam System pr escure ar e analyzed separately due to t he differences in aucunptions r equired f or a conservat ivo analysis. -

3.1.1 l'eak RCS Pressure Anal'/

/

sis 3.1.1.1 Hodalization Since the transient responsa of the turbine t rip event is the same for all loops, the single-loop model desctAbed in rection 3.2 of Reference 2 ,

in ut ilized f or thin analysis. The pressurizer tuodeling includes the use of the local conditions heat tr ansf er option f or t he vessel conduct ors .

3.1.1.2 Initial Ccnditions core Pawer Level liigh initial power level and a positive power uncertainty naximize the pritnary-t o-secondary power mismat ca upon tur bine trip. -

Prersuricer Prerrute Positive uncettainty is applied to the initial pressurizet preccute.

liigh init ial pressure r educes the initial tr.atgin to the overpressur e limit.

Pierruri;er Level High nitial level Ininimizes the initial volume of the pressurizer steam space, which maxituizen the transient primary pressure recponse.

Forr er ver rel Aver v'e Terrerum High initial temperature maximi 7es the primary coo ant stored energy, which maximizes the ttansient primary pressure response.

EN Flow Low initial flow mintmizes the primary-to-secondary heat transfer.

core Ovturr Flow Core bypass flow is not an ir:pottant parametet in this analysis.

3-1 a

. ]

rt e2.m Gener at er Level' High initial level minimites the initial volume of the steam generator steam space, which neaximizes the transient secondary pressure response.

Maximum secondary pressurization causes maximum secondary temperature response, which minimizes primary-to-secondary heat transfer, ruel 'remneraturo how fuel temperature, associated with high gap conductivity, maximizes tho t ransient heat t ransf er f rom the f uel to the coolant.

rt eren cenerat er Tube Pluacina A bounding high tube plugging value degrades primary-to-secondary heat transfer.

3.1.1.3 Boundary conditions Precrurizer rafety Valves l The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize the pressurizer pressure.

St e:Un Line Cafety Valves The steam line safety valves are modelei with lift, accumulation, and blowdown assumptions which maximize transient secondcry side pressure and minimize tranGient primary *to-BOCondary heat transfer.

3.1.1.4 . Control, Protection, and Safeguards System Modeling

-Reactor Trio The pertinent reactor t rip functions are_ the overtemperature AT (OTAT),

ocorpower-AT (OPAT), and pressurizer high pressure.

The response time of each of the two AT-trip functions is the Technical specification value. The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2.4.2 of Reference 2. In addition, the AT coefficients used in the analysis account for instrument uncertainties.

The response time-of-the pressurizer high pressure trip function is the Technical Spccification value. Since the pressure uncertainty is 1 accounted for in the initial pressurizer pressure, the pressurizer high

~

pressure reactor trip setpoint is the Technical Specification value.

Pressuricor Prersure control .

1 i

Pressurizer pressure control is in manual with sprays and PORVs disablea l

-in order to maximize primary pressure.

Pressuricer Level control )

Pressurizer level control is in automatic in order to elevate primary .

l pressure. Charging / letdown has negligible impact. l l

i 3-2

. _ _ . - , _ . ~ . . _ . - . _ _ . _ . _ - - . , ~ . _ . _ _ _ . _ . _ . . _ . _ . _ . _ . . _ . _

St erline I'CI"ZF an d Ccnd'mr " t e rn IM"

Secondary steam tolitt via the st eam line PORVs and the condenset s t (-am duttp is unavailable i ti older to maximize seccridary side pr essulization and minimize transient primary-to-secondary heat transfer.

r t e. + Ccnerntel 'r".l Cr ntid Feedwater is isolated upon turbine crip. The addition of subcooled f eedwat er would tend t o rubcool the watet in the st eado gener at or , and 1 educe secondaty side ptescute.

Ppd centrol

!Ja ctedit is taken for the operation of the Roa tont rol ryst em.

Following t ut bine t r ip, the turbine in pulse chattber presrure is rapidly reduced. The cortesponding reduction in the Rod Control System ref erence temperature would lead to cont t ol rod insertion, which would lessen the severity of the t ransient.

Auxiliary Feeint er Auxiliary feedwater la dicabled. The addition of subcoaled auxiliary f eedwater would tend to subcool t he water in the steam geretator, and reduce SeCondSry Side preEDute.

3.1.2 Peak Main Steam Systern Pressure Analysis 3.1.2.1 tJodalization Since t he transient tecponse of the tutbine trip event is the same for all loops, the 31raled oop taodel described in Secticr, 3.2 of Reference 2 is utilized for this analysis. The pressurizer modeling includes the use of the local conditions heat transfer option for tno vessel conductors.

3.1.2.2 Initial Conditions Ccr e Power Lovel High initial power level and a positive power uncertainty ruaxitntze the primary-to-secondary power micmatch upon turbine trip.

Precrurirei Precrute positive uncertainty is applied to the initial pressurizer pressure. As long as a high pressurizer pressure reactor trip is avoided, maxirnuin primary system preocute is conservative in order to delay reactor trip on OTAT.

Precrurizer Levd High initial level minimizes the initial volume of the pressurizer steam space, which maximizes the transient primary pressure response.

3-3

- ___ _ - _ - _ - _ _ - - - - _ - _ _ - _ _ _ _ _ _ - _ _ _ - - _ _ _ - - - _ - __ - _ _ _ _ . _ - ._ _- __ _----___ a

React or Vec 9eQgig_ Tem perat ure liigh initial teroperature maximizes the initial Main Steam System pres-sure and the prinary coolant stored energy.

Pr$. Flow liigh initial flow maximizes the primary-to-secondary heat transfer.

core Pyrarn Flow core bypass flow is not an important parameter in this analysis

team Generator Level liigh initial level minimizes the initial volume of the steam generator steam _ space, which maximizes the transient secondary pressure response. 3 l

Puol Temperatuig l Low fuel temperature, associated with high gap conductivity, naximizes the transient heat transfer from the fuel to the coolant.

t l'

fteam Generator Tube Pluanina Zero tube plugging is modeled to maximize primary-to-secondary heat j transfer.  !

3.1.2.3 Boundary conditicus Precrurirer rafety Valves 7

The pressurizer safety valves are modeled with lift, accumulation, and j blowdown assumptions which maximize the pressurizer pressure.

J

. Etesm Line fafetv Valven The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure.

3.1.2.4 Control, protection, and safeguards system Modeling Peactor Trip The pertinent reactor trip functions are the overtemperature AT (OTAT),

overpower AT (oPAT), and pressurizer high pressure.  :

The response time of each of the two AT trip functions is the Technical Specification value. The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2.4.2 of Reference 2. In addition, the AT.

coef ficients used in the analysis account f or instrument uncertainties.

The response time of the pressurizer high pressure trip function is the Tachnical Specification value. The pressurizer high pressure reactor trip setpoint is the Technical Specification value plus-an allowance which bounds the instrument uncertainty.

~

Pressuricer Pressure control pressurizer pressure control is in automatic with sprays and PORVs 3-4 u_ .

enabled in or der to I t event a high pressuriter pressure teactor trip actuation priot t o OTAT t r ip act uat ion.

PrcrruiliL1_ Level Centiel Pr essurl er level conct ol is in automatic in order t o c l evat e p r i!M ry pt m oute. Cha!gir.g/letd wn has negligiblG impact 9t e n Line ICEYr anu Ccndqnrer ';ter !*rp Secondary steam r elief via t he steftm line IOhVb and condenser 9 tears dunp ic uttavailable in ordet t o maximize second it y cide pressurizatiori.

Etg im Gene 1ater Level Cont rol reedwater is isolat ed upon turbine tr ip. The addition of subcooled

f eedwater would tend to cubcotl the watet in the steam generator, arid teduco secondary sido pressure. -

Pod ConQpl 11 0 credit is taken f or the operation of the Rod Control Syst em.

Following turbine trap, the turbine irapulse chamber pressure is rapidly teduced. The cor responding reduct ion in tne Rod Cont rol System s r ef erence temperatut e would lead to control rod inser tion, which would /

lessein the severity of the transient. ,

lg iliary Feedwater Auxiliary feedwater is dicabled. The addition of subcooled auxiliary 1 eedwat et would t end t o subcool the wat et in the steam generator, and reduce secundary side pressure.

3.. Lorr of 11cn-15e r c e ncy AC Power Tr The Station Auxiliaries A loss of non-emergency AC power causes the power supply to all busses not powered by the emergency d bsel generators to be lost. This leads

~

to t ho trip of both the nain f eedwater pumps and the reactor coolant pumps. A primary system h atup encues, due to both the coastdown of the reactor coolant pumps and the losc of main feedwater heat removal. As a result of this heatup, the primary concerns for this event are DlIB and ,

primary and secondary system overpressurization.

This transient differs from the complete loss of flow transient only in the timing of the insettion of the control rods. Both transients presume reactot coolant pump and feodweter purrp trip as the initiating events.

In the loss of flow event, the reactor trips when the reactor coolant pump bus undervoltage setpoint is reached and the rods begin to fall into the core after an instrumentation delay. In the loss of AC power transient, the control rods begin to f all immediately due to the loss of gripper coil voltago. Therefore, the transient core power response and consequently the DtJBR result is bounded by the loss of flow event.

Similatly, the primary system temperature increase and, therefore, the peak primary system pressure is also bounded by the loss of flow event.,

3-5 ,

1

.. . . U

Secondary side pressure does not rise significantly until the turbine trip occurs and steam flow is terminated. The magnitude of this pressure increase is largely determined by in amount of heat transfer 1ed from the primary system to the secondary once the pressute increara has begun. For this ever.t the reactor trip occurs prict to the turbine trip, such that the ptimary system heat generation is rapidly decreasing as_ secondary side presr.ure is increasing. Therefore, _the peak secondary pressure recult is bounded by the turbine trip event, in which the reactor trip occurs well after the turbine trip.

Based on the above qualitativo evaluation, e quantitative analysis of I this transient is not required. Should a reanalysis become necessary, eit her due to plant changes, modeling changes, or other changes which

-invalidate any of the above arguments, the analytical methodology employed would be as follows. 1 1

i Peak RCS pressure, peak Main Steam System pressure and core cooling capability are each analyzed separately due to the differences in 4 ascumptions required for a conservative analysis. The core cooling l capability analysis demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNBR remains above the 95/95 DNER ]

limit based on acceptable correlations. The minimum DMBR is determined l using the Statistical Core Design Methodology.

I l

3.2.1 Peak RCS Pressure Analysis i

i

? 2.1.1 Moda11:ation  ;

i Since the transient response of the loss of offsite power ovent is the l same for all loops, the single-loop model described in Section 3.2 of 1 Reference 2 is utilized for this analysis. The pressurizer modeling includes the use ot the local conditions heat transfer option for the vessel conductors.

> 3.2.1.2 Initial conditions I l

core Power Lovel High initial power level and a positive power uncertainty maximize the primary-to-secondary power mismatch, a

precsurirq Prewire Positive instrument uncertainty is applied to the initial pressurizer pressure. High initial pressure reduces the initial margin to the overpresourc limit.

Precouritor-Level High initAal level-minimizes the initial volume of the pressurizer steam space, which maximizes the ttansient primary pressure response.

3-6

Im;1 Yeru.1 Avei ro T( ~ ; ei st u r e liigh ini t ial t ettpet atut e ruxin t res t he initial prit:,aty coolant stoted 4 ttel gy , whi c h 11aXilhi Z es t he t i dnsient flimaty ptersure response, Ecr new Low initial flow degrades the prinaty-t omecondaty heat transtet.

cot e t":v a c c V)<?

tote /pess 110s is not an in mrtant pat amet er in this analysic.

Ttg e _generstor Le el Init ial st eam getwt at or level is riot an inpor tant par a uct er in this analyris.

Iuel i racrature Low f uel t empet it ur e, associated with high gap conductivity, maximizes the transient heat transfer from the fuel to the coolant.

rt er Gener au;1 Tub Pluuging A bounding high t ube plugging value degr ades pr imary-t o-secondary boat tranufer.

3.2.1.3 boundary conditions EP et,eration is11 four reactot coolant pumps are tripled at the initiation of the t ransient. . The punp model is adjusted such that the resulting coast down flow is conservative with respect t o the flow coastdown test data.

Pierrurizer rnie*" Valven The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumpt ions which maximize the pressurizer pressure.

e rt em Line nfetv valven The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize trannient secondary side pressure and minimize transient primary-to-secondary heat transfer.

3.2.1.4 control, Protection, and Safeguards System Modeling.

Eenctor Trii, The insertion of all control and shutdown banks occurs when the power is lost to the conttol tod drive mechanism.

Piereurizer Piorrure control Pressurizer pressure control is in manual with sprays and PORVs disabled in order to maximize primary pressure.

Plecturizer Level centrql Pressuriter level control as in automatic in orc'r to maximize primary pressure. Charging / letdown has negligible impact.

3-7

__ _ . _ _ _ _ _ _ _ _ _. J

]

. }

tpp Line roFVr and condenser rt e am Dm n i secondary steam relief via the steam 1)ne PORVs and the condenser cteam dump is unavailable in older to max 1n.1

e secondary side pres surization #

ar.d minind

e t r ansient pr imary-t o-secondary heat transfer.

i j k3xil!ary Feedwa W l Auxiliary f eedwater act uation occurs on the loss of offsite power after l an alpropriate TechnAcal Specification response time delay. A purgo I volume of hot main feedwater is assumed to be delivered prior t o tho l cold APW water reaching the steam ';enerat ors. In older to minimize the l post-ttip steam generator heat removal, the minituum auxiliary f eedwat er l flow is assumed.

i

Turb!ne Trit, j Turnine trip occurs on the loss of offsite power. i

, i t

l 3.2.2 Peak Main Steam system Presouro Analysis i

- 6 5 ,

. 3 . 2 , 2 .1. .Noda11:ution

• l Since the transient response of the loss of offsito power event is the -

j' same for all loops, the single-loop model described in Section 3.2 of

Reference 2 is utill

ed for this analysia. The pressurizet modeling '

l includes the use of the local conditions heat transfer option for the vessol conductors.

3.2.2.2 Initial conditions l Core Power Level. ,

! liigh -initial power level and a positive power uncertainty maximize the primary-to-secondary heat transfer.

I precruriter Pressure Pressurizer pressuro is not an important parameter in this analysis, i

Pierruriter LOVel since initial level primarily affects the transient primary pressure response, it is not an important parameter in this analysis.

Reacter verrel hveraae Tcm erature i liigh initial- temperature maximizes the initial Main Steam System pres- ,

sure and the prinary coolant stored energy.

Ecr Flow High initial flow maxinii:'es the prt": ry-to-secondary heat transfer.  !

Core Pvt > ass Flow Cote' bypass flow is not an important parameter in this analysis.

1; 3-8 E.

man Genent et Level High Initial level minimizes the init j al volutte of I he Bt ea n 90 net at or e t men r pa c e , which ::,sviri zes t he transient secondary precsute response.

Fuel Ter@ gat uig Low Iuel tempetatute, associat ed with high gap conductivaty tuximites the transient heat transfeI f rom t he fuel to the coolant.

?:t e.un O( nei at et T@e P1urina in older t o Inixirnize ju lmaly-t o-Locondar y heat tranclet, no tube plugging is raodeled.

3.2.2.3 houndaly Conditionc El.P Operatien All tour teactot coolant pumps tr ip on undervolt ago at the initiation of the loss of of f sit e power . The pump model in adjusted such that the teaulting coactdown flow 10 conser vat ive with respect to the 11cw coastdown test data. ,

herruilzer r;detv Valver The pr essurizer saf ety valvec are inodeled with lif t, accumulation, and blowdown assumptions which maximize pressu iter pressure.

St eMn line rafety Valvg The steam line caf ety valves are rnodeled with lif t, accumulation, and blowdown assumpt iono which maximi ze t ransient secondary side pr essur e.

3.2.2.4 Control, Proteccion, and Safeguards System Modelir'g Perter Trip C The insert ion of all contre 1 and shutdown banks occurs when tre power is -

lost to the control rod drive mechanism.

Pi mruricer Pi er r o rn control The operation of the prescurizer pressute control cyctem is not important in this analysis, h es ruricer Level cont rd The operation of t.he pressurizer level control system is not irnpot tant in thio analysis.

f,Leam Lino PORVr ani Condenrer Cteam h n Secondaty steam relief via the steam line PORVs and condenser steam dump is unavailable in order r.o maximize secondary side pressurization.

3-9

. . . . .. --_ _ _ _ . - __- _ ____ _ _ - - - _-_- _ _ _ _ - _ ___ _ A

- . ,- -.-.1 .--

Aux 111erv reedwat er i: Auxiliary f eedwater actuation occur s. on thu loss of of f site power af ter the appropriate Technical specificat ion recponse time delay. A purge volume of hot main feedwater ir assumed to be delavered prior to the cold AFW Water reaching the steam generators. In order to minimize the post-t rip st eam generator heat removal, the minirnum ata:iliary f eedwater flow is assumed.

Tuibitu;,T1ip Turbine t rip occurs on the loss of of f site power.

I 3.2.3 Cole Cooling Capability Analysis

{

3.2.3.1 Hodalization -

Since the transient zesponse of the loss of offsite power event is the

• same for all loops, the single-loop model described in Section 3.2 of Peterence 2 is utilized for this analysis. hie pressurizer 4.odeling includes the use of the local condit ions heat transf er option for the i vessel conductors.

3.2.3.2 Initial Conditions g re Power Level High initial powar level muximizes the primary system heat flux. The uncertainty in tois parameter is accounted f or in the Statistical Core Design-Methodology.

11g:;p,orizer Precrure Nominal full power pressuriter pressure is assumed. The uncerta:nty in this parameter is accounted for in the statistical Core Design

, tiethodol ogy.

Pressuriner Level Low initial level increases the volu.ne of the pressu:izer steam space ,

which minimizes the pressure increase resulting from the insurge. 1 Reactr,g Vensel Averaae Temrerature '

Nominal full power vessel average temperature is assumed. The uncertainty Ari this parar;.eter is accounted for in the Statistical Core -

Design Methodology. . t RCS Flow i i

Technical Specification minimum' measured Reactor Coolant System flow is

- assumed. .The uncertainty in this parameter is accounted ter in the ,

Statistical Core Design Methodology.

Core Bvrsass Plow The nominal calculated flow is assumed, with the flow uncertainty accounted for in-the Statistical Core Design Methodology, 3-10 u

& -, .. - - ,,,.--..m...-... ~ . . , - - , - _ ~ _ . . . . _ . _ , _ . . . , . - . . . - _ . . _ . , . - . . - , , _ . . _ . . _ _ . . _ - . . . . - _ _ _ . - . . . _ _ _ _ . ,

ne!" ceneiatoi t evi;l ,

Init ial stearn genet atol level in not an i.y>ortant pal anet er in this analysis, r fuel T4m m ture I A high init ial tenpelatur e is assumed t o minimize the gap condactivity calculated foi steady-ctate condi t i ora and used for the subsequcnt

> transient. A low cap conductivity minirnizes tho t ansient c har.g e in fuel rod sutface heat flux associated with a powel declease This. reakes t he power deC2 ease leEG seVete and is therefore consolvative for DNPR evaluation.

Steam Generator Tubu Plunaing scearn generatos tube pluggit.;; is not an it:por tant parameter in t his analysis. -

3.2.3.3 Doundary Condi". ions F_"r t e r Coolant Pumr All teactor coolant pumps are assumed t o trip on undervoltage et the initiation of the loss of offsito power. The purnp model is adjust ed such that the resulting coastdown flow is conservative wit. renrect to the Ilow coaatdown test data.

Dov Het

• End-of-cycle decay heat, baced upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is employed.

St eam Line Cafety Yalver The main steaN code safety v 1ves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.

3.2.3.4 Control, Protection, and Safeguarda System Mode]'ng Eert or Trip The insertion of all control 6nd shutdown banks occuta, when the power is lost to the control rod drive mechanism.

E m mul1 et Prcecure centroi Pressui1:er sprays aad FORVs are assumed to be operable in order !o minimize the system pressure throughout the transient.

Pressuri-er Level control Pressuri er heaters are assumed to be inoperable so thn Reactor Coolant System pressure is minitr' zed. Charging / letdown has negligible impact.

Eteam line PORVr and Conienser rt eam Lum Secondary steam relief via the stesm line PORVs and the condenser steam dump is unavailable in order to Inaximize secondary side pressurination and minimize transient primary-to-secondary heat. transfer.

3-11

-- _ - _ . __ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ - _ - _ _ _ _ _ _ _ _ - __ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ _ _ _ _ _ --_-__a

i Auxiliary Feedwater Auxiliary.feedwater actuati,n occurs on the loss of offsite power after the appropriate Technical Specification response time delay. A purge )

volume of hot nain feedwater is assumed to be deliversd prior to the cold APW water reaching the steam generators. In orcer to minimize the post-trip stwam generator heat removal, the riunimum auxiliary feedwater flow is assumed.

Pirbir.e Trin Turbine trip occurs on the loss of offsite power.

i 3.3 Loss Of-Mormal Feedwater I

A lows of normal f eedwater flow event could result due to the failure of l both of the msin feedwater pumps or a malfunction of the feedwater I control valves. A primary system heatup ensues due to the degradation-of the secondary heat sink. As a result of this heatup, the primary concerns f or this event are DNB and primary and secondary system overpressurization.

The loss of nornal- f eedwater transient is bounded by the turbine trip transient.- Both transients involve a mismatch between primary heat source and secondary heat sink, but the mismatch is greater for the tutbine trip. This is mainly due to the reactor trip and turbine trip occurring simultaneouoly for the loss of feedwater event,-whereas 1 reactor trip lags the turbine trip during the turbine trip transient. i

-Based on the above qualitative evaluation, a quantitative ana.ysis of this transient is not required. Should a reanalysis bect . necessary, either due to plant changes, modeling changes, or other changes which invalidate any of the above arguments, the analytical methodology employed would'be as follows.

Peak RCS presacre, peak Main Steam System pressure and core cooling capability are each analyzed separately due to the differences in

-assumptions required for a conservative analysis. The core cooling capability analyais demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined a using-the Statistical Core Design Methodology.

-3.3.1 Peak RCS Pressure Analysis 3.3.1.1 Nodalization Since the transient response of the loss of normal feedwater event is the same for all loops, the single-loop r..odel described in Section 3.2 of Reference 2 is utilized for this analysis. The pressurizer modeling includes-the use of the local conditions heat transfer option for the vessel conductors.

3-12

~

1

[ 3.3.1.2 Initial Conditions L

core Power Level liigh initial power level and a positive power uncer t ainty maxau te the piirnary-to-secondary power mismat ch.

Preccurizer Prerrure Positivo instrument uncettainty is applied to the initial pressutiter pressure. Ifigh initial prescure reduces the initial margin to the overpressure limit.

h g nitirer Level

)

High initial level minimizes the initial volumc of the pressurizer steam space, which maximizes the trar'sient primary pressure responso.

Reacter vercel Averaac Tgrerat ure i lidgh init ial t emperature maximizes the initial crimary coolant stored energy, which triaximites the transient pritrary pressure response.

ECO Ploy 1 Low initial flow degrades the pritrary-to-secondary heat trensfer.

A cele tiyearc Flow cote bypass flow is not an ittportant parameter in this analysis.

Eteam Generater Level

!!igh initial level in all steam generators delays reactor trip on low-low level and maximizes the heatup of the primary system .

3 Fuel Temr+ r at u r e Low fuel temperature, associated with high gap conductivity, maximizes the transient heat transfer from the tuel to the coolant.

rteam cenerator Tube Pluacina A bounding high tube plugging value degrades primary-to-secondary heat trancier.

3.3.1.3 Boundary conditions Precruri*er rafetv Valver The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize the pressuri:er pressure.

Eteim Line Safetv Valven The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure

'and minimize transient primary-to-secondary heat transfer.

Decay Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncert ainty, is employed.

3-13

__ _. -________________0

3.3.1.4 control, Protection, and f:sf eguards System !!odeling peactor Trin React or trip occurs when the low-low level setpoint is reached in t he steam generator.

,t3ercurizer Precrute control Pr essurizer pr essure conttol is in raanual wit n sprays and p0RVs disabled in order to maxirnize prin ary pressure.

h errurirer Level cent i el Pressurizer level control is in automatic in <cder to maximize primary pressure. Charging / letdown has negligible impact, steam Line P3ve and cendoncer st em r>um Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and miniinize t Mnsient prirnary-to-secondary heat transf er.

Eod control 110 credit in taken for the operation of the Rod Control cysten for this transient, which results in an increase in RCS ternporature. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt tc mairst.ain RCS temperature at its nominal value.

72rbine centrol The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.

Aux 11iary.Feedwater Auxiliary feedwater actuation occurs on low-low steam generator level after the appropriato Technical Specification response time delay. A purge volume of hot main feedwater is assumed to be delivered prior to the cold AFW water reaching the steam generators. In cider to minimize the post-trip steam generator neat removal, the minircum auxiliary feedwater flow is assumed.

3.3.2 peak Main Steam System pressure Analysis 3,3.2.1 14odalization Since the transient = response of the loss of normal feedwater event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2~1r utilized for this analysis. The pressurizer modeling includes the use of the local conditions heat transfer option for the vessel conduct ors.

3-14

_ _ _ ,_.u.. _ _ , ._ - _ _ _ . _ .._ . _ - _ _ . _

3.3.2.2 Tuitial Conditions core In ' 'e r Level liigh init ial power level and a posit ;ve power uncertainty naximize t he ptin.aty-to-secondary powet mismatch.

hertutirei h et un Pteucut'iner pressure is not an in:por t ant pat tu:.eter in this analysis.

h e r on,1;gr Level liessurizer level is not att inpottant parameter in this analysis.

Iieart er Ver rel Averaae Terix raturf liigh initial terperature maximizes the initial Main Steam System pres-sut e and t he primary coolant stored energy. -

PCC Flo;i liigh initial flow Inaximizes the pliinat y-to-secondary heat transfer.

Core Dynarr Pley <

Cote bypass flow is not an inportant pat amet er in thr u analysis ,

rteam Generator Level High initial level in all steam generators delayn reactor trip cn low-low level. Also, high level minimizes the initial volume of the steam generator steam space, which maximizes the transient secondary pressute response.

Fuel Temperatun Low f uel temperat.ute, associated with high gap conductivity maximizes the transient heat transfer from the f uel t o the coolant .

rt eam Generat er Tube Pluaninq Zer o tube plugging is modeled to rnaximize primary-to-cecondary heat -

transfer.

3.3.2.3 Boundary cont.itions Pr er surizer rafetv Unives The pr essurizer saf ety valves are Inodeled with lif t, accumulation, and blowdown assumptions which maximize the pressurizer pressure.

Steam Line saferv valvec The steam line se.fety valves are modeled with lift, accumulation, and blowdown assumptionn which maximize transient secondary side pressure.

Decav Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-sigma uncertainty, is emp'oyed.

3-15 l a

b a.3.2.4 Control, Protection, and Safeguards System Modeling Fonctor Trig Reactor-trip occurs when the low-low level setpoint is reached in the steam generator.

Pr e c r u r i ze r Prerruro control The results of this transient are not sensitive to the operation of preSFurl:er preBSure Control as long aD the pressure is controlled to within the range that avoids protection or safeguards actuation.

Prossurizer Level control I The results of this transient are not sensitive to the operation of pressurizer level control as long as the level is.kept within the rango that avoids protection or safeguards actuation. .,

l Tteam Line PORVs and condenter Steam _ Dump Secondary steam relief via the steam line PORVs and condenser steam dump is unavailable in order to maxirt.ize the transient secondary side pressuriration.

Pod Control No credit is taken for the operation of the Rod Control System for this i transient, which results in an increase in RCS temperature. Wit h the Rod-Control System in automatic, the contr01 rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RC; temperature at its nominal value.

Turbine centrol The tutbine is modeled in the load control mode, which is described in

. Section 3.2.5.1 of Reference 2.

l l

Auvillary Feedwat er.  ;

l Auxiliary feodwater actuation occurs on low-low steam generator level after the appropriate: Technical Specification response-time delay A purge volume of hot main feedwatec is assumed to be delivered prior to the cold AFW water reaching the steam generators. In order to minimize the post-trip stoam generator heat removal, the minimum auxiliary l

feedwater flow--is assumed.

3.3.3 Core Cooling capability Analysis 3.3.3.1. Nodalization

'Since the transient response of the loss of normal feedwater event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis. The pressurizer modeling-includes the use.of the local conditions heat transfer option for the vessel conductors.

3-16

,.. _ _ _ _ _ . . _ _ _ _ _ ._ _.. _ ._. _ _ _ _ -- -_._.s . - _

j 3.3.3.2 Initial Conditions 4

21e Power Level High initial power level maximizes the primary system heat flux. The uncert.ainty in this paramet er is acccunted f or in the Statistic 6' Core Dosign Methodology.

Frertur12o1 Pre gulfe flominal full power pressurizer pressure is assumed. The uncertainty in this parameter 17 accounted for in the Statistical Core Design Met hodology .

Eternurizer Level Low initial lovel increases the volume of the pressurizet steam space which minimizes the pressure increase resulting from the insurge.

Reactor Vorrel Averace Terrerature Nominal full power vessel average temperature is assumed. The uncertainty in this parameter is accounted for in the statistical core Design Methodology.

RCE Flow Minimum measured Reactor Coolant system flow is assumed. The uncertainty in this parameter is accounted f or in the Statistical Core Design Mathodology.

Core Pyparc FJ ow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Core Design Methodology.

Eteam Generator Tube Pluaaina A bounding high tube plugging level impairs the ability of the secondary side to remove primary side heat.

Fuel Temiser3ture A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient. A low gap conduvLivP" minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNBR evaluation, steam Generator Level High initial level in all steam generators delays reactor trip on low-low level and extends the primary system heatup.

3.3.3.3 Doundary Conditions ste n Line saferv valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maxi.nize secondary side pressure and minimito primary-to-secondary heat transfer.

3-17

-_ _ _ _ _ - _______ _ ___- __ _ __ _ a

Decav_ Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a two-signa uncertainty, ir employed.

3.3.3.4 Control, Protection, and Safeguards System Modeling Reactor Trip Reactor trip occurs when the low-low level setpoint is reached in the steam generator.

] l Precrurirer Pre 9sure centrol i Pressurizer sprays and PORVs are assumed to be operable in order to minimize the system pressure throughout the transient.

1 frorruriner Level control Ho credit 1: taken for pressurizer heater operation so that Reactor Ccolant System pressure is minimized. Charging / letdown has negligible impact.

-I

. Steam Line PDPVs and condenser steam Dump A

Secondary steam relief via the steam line PORVs and the condenser steam j I

dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transf er.

.End control ths credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod Control-System in automatic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value.

Turbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1-of Reference 2.

Auviliarv Feedwater Auxiliary.feedwater actuation occurs on low-low steam generator level after the appropriate Technical Specification respon e time delay. A purge .olume'of hot main feedwater is assumed to be delivered prior to the cold AFW water reaching the steam generators. In order to minimize

.the post-trip steam generator heat removal, the minimum auxiliary feedwater flow is assumed.

Turbine Trio Turbine trip occurs on reactor trip.

3.4 Feedwater ?vstem Pit,e Preak The feedwater system pipe break event postulates a rupture of the Main Feedwater System piping just upstream of the steam generator (downstream of the tinal'feedline check valve).

- Following the blowdown of the faulted' generator, there is a mismatch between the heat generation in 3-18

a. . .

) the teacter and t he recorrlary side heat r m val late gue t o t he mismat ch, the primaly concern for this transient is the cat ability t o

+

effectively cool the 2eactor core Adequate chcrt term and long term core cooling capability .re analyzed separately due to t he diffelence: In assumpticns required for a concer vat ive ar.alysiu . The short tern core cooling capability analyris danonst rat es t hat fuel cladding int egrity is maint ained 1.y ensul ing t hat t) minimum EJ!i h r emainc above the 95/ 9 3 DI:LR limit based on accept al'1.t c t '_ elat j ons . The minimur.. DI!LP is det ermine? " sing the statistical core Dm:ign Met hodology The long term cote coolang capability analycis demonctrates that no hot ieg boiling occurs.

3.4.1 Sholt Term Cole Cooling Capability --

The D!iB analysis for thic transient is modeled as a complete los- of coolant. flow event. initiated from an off-norma'. condition. Tha loss or flow is assumed to occur coincident with the OTAT reactor t rip causcci by the feedline break heatup.

3.4.1.I !Joda1iration r;i nce the complete loss of flow trancient is cyn.m. ical wit h respect to the four reactor coolant loops, a single-loop model (Reference 2, section 3.2) is utilized for this analysis.

3.4.1.0 Initial Cenditions core P^wer LM High initial power level Inaximl:es the primary systelu heat flux. The "

uncettainty in this parameter is accounted for in the Statistical Core Design Methodology.

Prenuricer Presrure IJominal full power pressurizer pressute is assumed. The uncertainty in this parame;er is accounted for in the Statistical Core Design Methodology.

Prermarlzei Le"el Low initial level increases the volume of the pressurl:er steam space which minimites the pressure increase resulting from the insurge.

Pe rt er Ver rel Aver ge Terreraturg

!Jominal f ull power vessel average temperature is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.

3-19

. _ _ ___ _ _ . .-- -J

1

. J Rcr Flow q Minimum near.uted Reactor Coolant System flow is assumed. The

]

uncertainty in t his parameter is accounted f or in the Statistical Core Design Methodology.

l core Ierasr Flow /

The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical core Design Methodology.

Fuel Tennerat ure l A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the-subsequent transient. A low gap conductivity minimizes the transient change in ,

fuel rod surface heat flux associated with a power decrease. This makec the power decrease less severe and is therefore conservative for DMBR ,

evaluation. '

rteam cenerator Level Initia11 steam generator level is not an important parameter in this '

analysis. j rream caerator Tube Pluaaine For transients of such short duration, steam generator tube plugging does not have an effect on the transient results, i

I 1

l 3.4.1.3 Boundary Conditions '

Steam Line Fafetv Valvos The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.

3.4.1.4~ Control', Protection, and Safeguards System Modeling Enactor Trip.

Reactor trip occurs on overtemperature AT following the heatup due to the heat transfer mismatch. Earlier trips on high containment pressure safety injection and low-low steam genecator level are not credited in order to naximize the primary system heatup.

Pressurizer Preccure control

)

Pressuriser sprays and PORVs are assumed to be operable in order to-minimize the system pres 1ure throughout the transient. 1 Pressurizer Level control  ;

-1

-Pressurizer. heaters are assumed to be inoperable so that Reactor Coolant j Fystem pressure is minimized. Charging / letdown has negligible impact.

i 3-20 c.__.......,

, rt er. Line 00Rve and condenrer rteam rene Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.

( Eod control No credit is taken for the operation of the Rod ContI: 1 System for this

, transient, which results in an increase in RCS t amper i.tu r e . With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as they are insertei in an attemR to maintain RCS temperatute at its nominal value.

Turbine control ~

The turbine is modeled in the load control mode, wrich is described in Section 3.1.5.1 of Feference 2. -

Auxiliary Feedwat er AFW flow would be credited when the safety analysis value of the low-low steam generator levvl setpoint is reached. However, the parameter of inter est for this transient has reached its limi:'ing value before the -

appropriate Technical Specification response tire delay has elaped.

Therefore, no AFW is actually delivered to the uteam generators.

T2rbine Trig ,

The reactor trip leads to a subsequent turbine trip. t 3.4.2 Long Term Core Cooling capability 3.4.2.1 Nodalization Due to the asymmetry of the auxiliary feedwater flow boundary condition in the f eedline break tr ancier.". , a three-locp model (Reference 2,Section -

3,2), with two single loops ari one double loop, is utilized for this Snalyris.

3.4.2.2 Initial Conditions core Power Level High initial power level and a poaitive power uncertainty maximize the primary sy" eru heat load.

Erne ruri rgr_?m sur e Low initial pressure causes a corresponding decrease in the hot leg saturation temperature, which minimizes the margin to hot leg boiling and is conc.:rvat.tve for demonstrating long term core cooling.

Prescurizer Level Low initial level increases the volume of the pressurizer steam space which ninimizes the pressure increase resulting f rom the insurge.

3-21

_______ __ _ _ _ _ _ _ - - - - - - - - - - - - - - - __ -J

. ptgetor vesrel Averaae Temnerafure

. iiign initial temperature increases the stored energy in the primary system which must be removed by the degraded secondary side.

=Ecs Flow Low initial flow degrades the primary-to-secondary heat transfer.

Core Pyoars'Ficw Core bypass flow is not cn important parameter in this analysis.

Steam Generator Level Low initial level in all steam gene: Ators decreases the long-term capabilityf of _ the secondary system to remove primary system heat ,

Funi Temnerature A conservatively high initial 1 fuel temperaturr a: ,umed in order to l maximize the amount of stored energy that mus- '. temoved.

steam cenerator Tube pluacina-A boundir.g high tube plugging level impairs the ability of the secondary side to remove primary side heat.

3.4.2.3 Boundary ( tions

~

)

Break M6de11na J The feedline break is modeled as a double-ended rupture of the main )

feedwater line=just upstream of the steam generator (downstream of the I check valve). A bounding flow area of the break junction is assumed in order to maximize the break flowrate. The break flowrate is determined by the Henry ( subcor '.ed) and Moody (saturated) critical flow correlations.

l Rea'ctor Coolant Pumpg. '

The timing of the operator action to trip the reactor coolant pumps is

investigated in a sensitivity study. Based on the results of this sensitivity-study, an early pump trip time, with the correspondi - l natut..! circulation heatup, is. conservative. The RCPs are tripped at 15

'seccr4s, which.is assumed to precede the time at which the pu-- would be manually tripped on high-high containment prescare.

i Pressuri?cr-Safetv Valves  !

.The pressuriter safety. valves are modeled with lift, accumulation, and .

blowdown assumptions;which minimize pressurizer pressure.

l 1

SteEm Line Safety Valves The main steam code re:ety valves are modeled with lift, accumulation. j andiblowdc,n'assumptio.o which maximize secondary side pressure and l minimize primary-to-secondary heat transfer. l Decav Heat End-of-cycle decay heat, based upon the ANSI /ANS-5.1-1979 standard plus a-two-sigma uncertainty, is employed.

3-22

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3.4.2.4 Control, Protection, and Safeguards System Madeling

(

Peact or Tiin The reactor is tripped 10 seconds into the transient. This is assumed to be after the occurrence of safety injection actuation on high containment pressure.

_Pi_e s c ur i zeL_Pr e s r u t o centrol Since low Reactor Coolant System pressure is conservative and the blowdown pressure of a cycling sa'ety valve is much lower than for a cycling PORV, the PORVs are assun ed inoperable. T escurizer spray is assumed to be operable in order to minimize system pressure.

Precurizer Level control -_

Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System precsure is minimized. Charging / letdown has negligible impact.

Steam Line PORVc and Condenser Steam Duro Secondary steam relief via the steam line PORVs and the condenser steam 0 dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat tr~nsfer.

Rod Control 11 0 credit is taken for the operation of the Rod Control System for this trancient, which results in an increase in RCS temperature. With the Rod Control Eystem in automatic, the control rods would cause a negative teactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value.

Turbine centrol The turbine is modeled in the load control mode, which is described in Secticn 3.2.5.1 of Reference 2.

safety Iniection Safety injection actuation occurs at 10 seconds on high containment pressure. One-train minimum injection flow, as a function of RCS pressure, is assumed to minimize the delivery of cold SI water.

Injection is stopped when the emergency procedure SI termination '

criteria are met.

Auviliary Feedwater Auxiliary feedwater actuation occurs on safety injection actuation after the appropriate Technical Specification response time delay. A purge volume of hot water is assumed to be delivered prior to the cold AFW water reaching the steam generators. Operator action to isolate AFW flow to the faulted generator occurs at 120 seconds as a result of a sensitiv2ty study. In order to minimize the post-trip stearn generator heat removal, the minimum auxiliary feedwater flow is assumed.

3-23

-___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -. -- a

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'MsIV closure (

The' timing of the closure of~the main steam isolation valves-is the

- focus of a sensitivity study which shows that early MSIV closure, which

' initiates - the overheating phase of the transient, - is conservative. The valves are closed at':15' seconds, which is assumed to precede automatic closure on high-high containment pressure.

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4.0 DECREASE Ill REACTOR COOLAi!T SYSTEM FLOW RATE 4.1 Par t 131 Lese of Forced Rearter Coolant Flow A partial loss of forced reactor coolant flow can result ficai a rnechanical or electrical f ailure in a reactor coolant pump, or from _t fault in the power supply t o the pump. If the reactor is at power when such a fault occuts, this could result in DNB with subsequent fuel damage if the teactor is not tripped promptly. The necessary protection apsinet a partial loss of coolant flow is provided by the low reactor coolant flow reactor trip signal.

The acceptance criteria for this analysis are to ensure that there is adequate core cooling capability and that the pressure in the Reactor _

Coclant System remains below 110% of design pressure. The cote cooling capability analysis demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNER remains above the 95/?5 DNBR limit based on acceptable corrolations. The minimum D!ER is determined using the Statistical Core Design Methodology. The peak RCS pressure criterion is met through a comparison to the peak pressure results for the more limiting locked rotor transient. In Section 4.3 of thic report, the locked rotor event is shown to remain below 110% or the RCS design pressure.

4.1.1 Nodalization Thic non-symmetric transient is analyzed using a two-loop model, with a single loop for the tripped reactor coolant pump and an intact triple loop. The ptessurizer modeling includes the use of the local conditions heat transfer option for the vessel conductors.

4.1.2 Initial Conditions ffC,re Power Level High initial power level raaximizes the primary system heat flux. The uncertainty for this parameter is incorporated in the S atistical Core Design Methodology.

Preocuri;pr Prescum The nominal pressure corresponding to full power operation is assumed, with the uncertainty for thic parameter incorporated in the Statistical Core Design Methodology.

Preceurizer Level Low initial level increases the volume of the pressurizer steam space which minimizes the pressure increase resulting from the insurge.

Reactor vessel Averaae Temneratu m The nominal tempetature corresponding to f ull power ( eration is assumed, with the uncertainty for this parameter incorporated in the Statistical Core Design Methodology.

4 -- 1

___ _ _ _ _ _ _ _ _ - - - - _ _ _ _ _ _ _ _ - . _ - - - - -. --__ -._- _ .J

_ECS Flow-The Technical Specification minimum measured flow for power operation is

~

assumed sinceflow flow is conservative for DNBR evaluation. The uncertainty for this parameter is incorporated in the Statistical Core.

Design Methodology.  ;

Core Pvnass Flow The~ nominal' calculated flow is assumed, with the flow uncertainty accounted-for in the Statistical Core Design Methodology.

Steam Generator Level Initial steam generator' level is not an important parameter in this analysis.

Fuel Temnerature =

A high initial temperature is assumed to minimize the gap conductivity calculated for~ steady-state. conditions and used for the subsequent transient. A low gap conductivity minimizes the transient change in fuel-rod surface heat flux associated with a power decrease. This makes the power. decrease less severe and is therefore conservative for DNBR evaluation.

Eff n Generator Tube Plucains For. transients of such-short duration, steam generator tube plugging I does not have an effect on the transient results, 4.1.3 _ Boundary-Conditions RCP'Oneration-A' single reactor-coolant pump is assumed to trip. The other three reactor coolant pumps: remain operating for the duration of the-transient. The reactor coolant pump model is adjusted such that-the

-resulting pump coastdown is conservative with' respect-to the flow

-coastdown test-data. -

Steam Line' Safety Valves-The: main steam code safety' valves are modeled with; lift, accumulation, and? blowdown assumpti*ons which maximize secondary pressure;and minimize primary-to-secondary heat transfer.

4.'1. 4 L Control, Protection, and Safeguards System Modeliag Egaetor Trin

_l' A reactor 1 trip signal is generated when flow in the affected' loop falls below a setpoint which conservatively bounds the Technical Specification value. -A delay' time consistent with the Technical Specifications is assumed-between receipt of the low flow signal and the-initiation of control rod motion.

J 4-2 l l

l

j hecourizer Presrure control Pressurizer sprays and FORVs are assumed to be operablt in order to minimize the system pressure throughout the transient.

Prescurizer Level control Pressurizer heaters are assumed to be inoperable so that Reactor Ccolant System pressure is minimized. Charging / letdown has negligible mpact .

Ste n Line PORVo and condenrer Ster Durn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.

Steam Generator Level control The results of this transient are not sensitive to the mode of steam -

generator level control as long the level is kept within the range that avoids protection or safeguards actuation.

MFW Pumn Sneed control The recults of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generator level is kept within the range that avoids protection or safeguards actuation.

Rod control No credit is taken for the operation of the Rod Control System for this transient, which results 5 an increase in RCS temperature. With the Rod Control System in autA:.atic, the control rods would cause a negative reactivity addition as they are inserted in an attempt to maintain RCS temperature at its nominal value Turbine control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.

Auxiliarv Pe dwater AFW flow would be credited when the safety analysis value of the low-low steam generator level setpoint is reached. However, the parameter of interest for thir transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed.

Therefore, no AFW is actually delivered to the stearo generators.

Turbine Trio I The reactor trip leads to a subsequent turbine trip.

1.2 Corelete Lost Of Forced Reactor Coolant Flow A complete loss of forced reactor coolant flow would occur if all four reactor coolant pumps tripped due to either a common moue failure or a simultaneous loss of power to the pump motors. The Reactor Protection System {RPS) senses an undervoltage condition at the pumps and initiates a reactor trip. The decrease in core flow which occurs prior to reactor trip causes a heatup of the Reactor Coolant Sys. t em .

4-3

-. _ . _ _ _ - _ _ - ____- - -_ _ . _ _ _ _ D

1

.The acceptance criteria for this analysis are to ensure that there-is adequate core cooling capability and that the pressure in the Reactor Coolant SystemLremains below 110% of-design pressure. The core cooling-capability analysic. demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNBR remains above the 95/95 DNDR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology. The peak RCS pressure criterion is met through a comparison to the peak pressure results for  ;

the more limiting locked rotor transient. In Section 4.3 of this report =the locked-rotor event is shcwn no remain below 110% of the RCS design pressure.

4.2.1 Nodalization

\

Since tha complete loss of flow transient is symmetrical with respect to

-the four reactor coolant loops, a single-loop model-(Reference 2, section 3.2) is utilized for this analysis.

4.2.2 Initial Conditions

-Core Power Leggi

-High initial power level maximizes the primary system heat flux. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology, Pressurizer' Pressure

Nominal full power pressurizer pressure is assumed. The uncertainty in this parameter is accounted for-in the Statistical Core Design Methodology.

Pressurizer Level Low initial level increases:the. volume of the pressurizer steam space Ewhich minimites-the' pressure increase resulting from the insurge, Reacter Vessel Averace Temnerature Nominal. full power vessel: average temperature is assumed. The c uncertainty in'this parameter is accounted for in the Statistical. core Design Methodology.

.RCS Flow.

' Minimum measured Reactor Coolant System flow is assumed. The uncertainty-in this~ parameter is accounted for in the Statistical Core Design Methodology.

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'Eyrass Flew The nominal' calculated flow corresponding to full power operation is assumed, with-the flow uncertainty accounted for in the Statistical Core Design Methodology. ,

Steam Generator Level ~ l Initial steam generator level is not an important parameter in this E -analysis.

- 4-4

Fuel Temnerature A high initial temperatute is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient, A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNBR evaluation.

fream Generator Tube Pluaaing For transients of such short duration, steam generator tube plugging does not have an effect on the transient results.

4.2.3 Boundary Conditions _

E;P operation All four reactor coolant pumps are tripped at the initiation of the transient. The pump model is adjusted such that the resulting coastdown flow is conservative with respect to the flow coastdown test data, steam Line saterv Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximizc secondary side pressure and minimize primary-to-secondary heat transfer.

4.2.4 Control, Protection, and Safeguards System Modeling Reactor Trin Reactor trip occurs on reactor coolant pump undervoltage, after an appropriate instrumentation delay.

Precrurizer Preceure Control [

Pressurizer sprays and PORVs are ascumed to be operable in order to minimite the system pressure throughout the transient.

Prescuricer Level control Pressurizer heaters ace assumed to be inoperable so that Reactor Coolant System pressure is minimized. Charging / letdown has negligible impact.

Steam Line PORVs and Condenser Steam Dumo Secondary steam relief via the steam line PCRVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.

Steam Generator Level Control The results of this transient are not sensitive to the mode of steam generator level control as long the level is kept within the range that avoids protection or safeguards actuation.

4-5

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USW ren r r4 ed contysl The results of this transient are not sensitive to the mode of MFW pump speed control as long as the steam generat or level is kept within the range that avoids protection or safeguards actuation.

Pod cent rol No credit is taken for the operation of the Rod Control System for this transient, which results in an increase in RCS temperature. With the Rod control System in automatic, the control rods would cause a negative teactivity addition as they are inserted in an attempt to maintain RCS temperat ure at its nominal value.

Turbine contr21 The turbine is modeled in the loud control mode, which is described in Section 3.2.5.1 of Reference 2. i 1

Auxi 1i ary_feedwat e.

AFW flow would be credited when the safety analysis value of the lov-low I steam generator level setpoint is reached. However, the paramecer of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay has elapsed. l Therefore, no AFW is actually delivered to the steam generators.

7tibine Trip l The reactor trip leads to a subsequent turbine trip.

4.3 Ecarter coolant Purn Locked Peter The poetulated accident involves the instantaneous seizure of one reactor coolant pump rotor. Coolant flow in that loop is rapidly reduced, causing the Reactor Protection System (RPS) to initiate a reactor trir on low RCS loop flow. The mismatch between power generation and neat removal capacity due to the degraded flow condition causes a heatup of the primary system.

The acceptance criteria for this analysis are to onsure that there is adequate core cooling capability and that the pressure in the Reactor Coolant System remains below 120% of design pressure. Peak RCS pressure and core cooling capability are analyzed separately due to the differences in assumptions required for a conservative analysis. The core cocling capability analysis determines to what extent fuel cladding integrity is compromised by calculating the number of fuel rods that exceed the 95/95 DNBR limit based on acceptable correlations..

4.3.1 Peak RCS Pressure Analysis 4.3.1.1 Modalization Due to the asymmetry of the transient, a two-loop model (Reference 2, Section 3.2), with a faulted single loop and e intact triple loop, is utilized for this snalysis. The pressurizer modeling includes the use 4-6 m i

of the local conditions heat t rar.s ter opt ion f or the vessel conductcrs.

4.3.1.2 Initial conditions 1

fJg;ft_1l2yter Level High init ial power level and a positive powel uncertainty naximize the primaty system heat load.

Prvrouriner Pterrura liigh initial pressure yields a cmaller margin to overpressur ization.

Etesaurizer Lcvel -

High initial level decreases the volume of the pressuriner steam space which maximizes t he pv0ssure increase rcsulting from the insurge. ,

Reactor Vecre! Averaw Teer u ur e High initial temperature maximizes the initial primary coolant stored energy, which maximizes the transient primary pressure response.

RCE Flow Low initial flow minimizes the primat,/-to-secondary heat transfer.

Core Dyr,ane Flod High core bypass flow minimizes coolant flow through the core and exacerbates heatup.

Steam cenerator Level Inittal steam generator level is not an important parametet in this analysis.

Puel T.cr,erature Low tuel temperature, associated with high gap conductivity, manimizes ,

the transient heat transfer from the fuel to the coolant. -

Steam Generator Tube Phmirs For transients of such short duration, steam generator tube plugging does not have an offeet on the transient results.

4.3.1.3 boundary Conditions Peactor Conlant Pumq The rotor of the reactor coolant pump in the 'aulted loop is assumed to seize at the itii t iation of the transient. The remaining reactor coolant pumps trip on bus undervoltage fullowing the loss of offsite power.

Orf cit e Power offsite power is assumed to be lost coincident with the turbine trip.

Prersurizer snfety vaine The pressuript safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize pressurizer pressure.

4-7

-Steam Line-Saferv Valves

.The main steam codeLsafety valvos are modeled with lift, accumulation,

-and blowdown assumptionsLwhich naximize secondary side pressure and minimize primary-to-second&ry heat transfer.

14'.3.la4 ' Control, Protection, and Safeguards System Modeling Peactor Trip Reactor trip occurs on low Reactor Coolant System flow in the locked loop.

- h surizer Prese:ure Control

In order to maximize primary system pressure,-no credit is taken for pressurizer spray or PORV operation. .

.frescurizer Level control Pressurizer heaters are assumed to be operable in order to maximize Reactor Coolant System pressure resulting from the insurge/ level  !

increase. Charging / letdown has negligible impact.

Steam Line PORVs and' Condenser Steam Dumn Secondary steam relief via the steam line-PORVs and the condenser steam ,

dump is unavailable in order to maximize secondary side pressurization i and minimize transient 1 primary-to-secondary heat transf er.

jigam Generator Level Control ,

The results of this transient are not sensitive to the mode of steam

_ generator level contrcl as long the level is-kept within the range that u avoids protection or safeguards actuation.

I14FW ' Pumn Sneed Centrol LThe results_of_this transient are not sensitiva to the mode of MFW pump speed control as long as the steam generator level is kept within the range chat avoids-protection-cr safeguards actuation.

Rod control No ctedit is taken=fo' _the:-operation of the Rod Control System for this transient, which results in an increase in RCS tempera ;ure. . With the-Red Control System in automatic, the control rods would cause a negative reactivity addition as they are. inserted in an attempt to maintain RCS i

--temperature-at its nominal value.

Turbino control

~

.The turbine .is modeled in the load control mode, which is' described in ,

Section-2.2.5.1 of_ Reference 2. )

Auxiliary Feedwater

>J31 flow would be credited when the saf ety analysis -value of the low-low steam generator' level setroint is reached. Howevar, the parameter of

'interect for this transient has reached its limiting value before-the

! - -appropriate" Technical Specification response time delay has elapsed.

Therefore, no AFW is actually delivered to the steam generators.

4-8 '

d Turbine Tr g The t eactor trip leads to a subsequent turbine trip.

s 4.3.2 Core Cooling Capability Analysis 4.3.2.1 Nodalization Due to the asyrr:netry of the transient, a two-Joop model (Reference 2, Section 3.2), with a single (faulted) loop and a triple (intact) loop, is utilized for this ar21ysis. The pressurizer modeling includes the use of the local conditions heat transfer option for the vessel conductors.

4.3.2.2 Initial Conditions C

Coro Power Lovel High initial power level and a positive power uncertainty maximize the primary system heat load.

Pressurizer Pressure Low initial pressure yields a lower initial, and therefere transient, DNBR.

Eyssurizer Level ,

Low initial level increases the volume of *:he pressurizer ? team space which minimizes the pressure increase resulting from the insurge.

Fenctor Vessel Avernae Temeerature High initial temperature increases the stored energy in the primary system which must be removed and miriimizes the transient DNBR.

7 RCS Flow Low initial flow degrades the primary-to-secondary heat transfer and itinimizes the transient DNBR.

Core D/rass Flow High core bypass flow exacerbates heatup by minimizing coolant flow through tha core and minimizes the transient DNBR.

Oteam Generator Level ,

Initial steam generator level is not an important parameter in this analysis.

Fuel Temperature A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions and used for the subsequent transient. A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated witF a power decrease. This makes the power decrease less severe and is therefore conservative for DNBR evaluation.

4-9 l

- _ ___ _ - _ - - _ _ _- ___- - _ _ _ _ .________________-___ - ____________-___ ____________ - __ _O

Steam Generator Tube Plunaina For transients of such shortEduration, steam generator tube plugging

- does_not have an~effect on the transient results.

4.3.2.3 Boundary. conditions Reactor Coolant Pumns -

The rotor of the reactor coolant pump in the faulted loop is assumed to seize at the initiation of the transient. The remaining reactor coolant .

. pumps trip on bus undervolta;3 following the loss of offsite power. -

Cf f site Power I Offsite power is assumed to be lost coincident with the turbine trip.

Pressurizer Safety Valven The pressurizer safety valves are not challenged by this transient.

Steem Line Safety Valves The main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and j' minimize primary-to-secondary heat transfer.

4.3.2.4 Control, Protection, and Safeguards System Modeling Egaetor' Trio Reactor trip occurs on low Reactor Coolant System flow in the loop with -i the locked rotor. l l

l t' .Pressuri?er Pressure Contrgl Credit ~1s taken for both pressurizer spray and PORV operation in order to minimize primary system pressure.

Pressurizer Level Control Pressurizer heatere are assumed to be inoperable so that Reactor Coolant

. System pressure _is minimized. Charging / letdown has negligible impact.

< Steam Line PORVs and Condenser Steam D'imo Secondary steam relief via the steam line PORVs and the condenser steam dump is. unavailable in order to maxa ize secondary side pressurization and minimiz~e transient primary-to-secondary heat transf er.

Steam Cenorator Level Control

.The results of1this transient are not sensitive to the mode of steam generator level control as long the level is kept within the range that avoids protection or safeguards actuation.  !

.MFW Pumn Soeed Control- l

, The resultc of this transient are not sensitive to the mode of MFW pump 1 speed-control as-long as the steam generator level is kept within the range:that avoids protection or safeguards actuation.

4-10 L( i

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hd centrd 11 0 credit is taken for the operaticn of the Pod Cont rol Systee f or this transient, which result s in an increase in RCS temperatur.>. With the Rod Control System in automatic, the control rods would cauce a negative react. ivi t '/ addition as they are inserted in an attempt te maintain RCO temper atur e at its nominal value.

'"urbine Conti el The turbine is modeled in che load control mode, which is described in Section 3.2.5.1 of Reference 2.

Aue!11arv Feete m AFW flow would be credited when the sater v analysis value of the low-low st eam generator level setpoint is recached. However, the pa ranw t er of interest for this transient has reached its limiting value before th-appropriate Technical Specification response time delay has elapsed.

Therefore, no AFW is actually de'ivered to the steam generators.

Tarbine Tu n The reactor t.ip loads to a subsequent turbine trip.

4.3.2.5 other Assumptions y The peak clad temperature calculation employs the fuel conauction model as described in Section 4.2.2 of Reterence 1.

9 41 0~11 4

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5.0 REACTIVITY A'iE POWER DISTRIBUTION ANOMALIES 5.1 Uncont i t lled Pank Wi thdr awal Fy n a_f,;& iitical ei tc Nwn rtartun rdif, i.2n A malf unct ion of the Rod Control System can result in an uncontrolled 1 withdrawal of control rods. Beginc.ing frcm a low initial power typleal of Modes 2 and 3, the msul ng positive r eact ivity addition causes a 1 power excursion which is terminated by the high power range flux (low setpoint) RPS trip function. Since the initial condition requires as few as three teactor coolant pumps in operation, the minimum DNbR is of concern for peak transient power levels less than full power. The peak Reactor Coolant System pressure limit of 110% of design pressure is also of concern due to the mismatch between core power and the secondary heat _

sink during the power excursion. Peak RCS pressure and cote cooling capability are analyzed separately due to the differences in assumptions required for a conservetive analysis. The core cooling capability analysis demonttrates that fuel cladding integrity is maintained by ensuring that the minimum DNBP remains above the 95/95 DNBR limit based y on acceptable cor rela tions . The minimum DNER is determined using the Statistical Core Design Methodology.

e 5.1.1 Peak RCS Pressure Analysis 5 .1 -. l .1 Modalization The peak RCS pressute transient is analyzed with four reactar coolant pumps in c perat i on. Since all jnitial and boundary conditions are symn at ric , a single-loop model or any multi-loop nodalization is appropriate. The standard model (Reference 2, Section 3.2) is naed with -

one significant exception. Since this transient initiates at zero power, -

dnd since the duration of the trcnsient is "ery 3hort, the steam 8 gene 1ator secondary response is not important. Rather than using the standard steam generator secondary nodalization, a single secondary volume is used. The single volume uses the bubble rise option with the local-conditions heat transfer model applied to the steam generator tube conductor 4 With this mcdeling approach the initial condition of zero power can be obtainea, ard the primary-to-secondary heat transfer that occurs followit the p wer excursion can be simulated. The pressurizer modeling includes the use of the local conditions heat transfer option for the vessel conductors.

9.1.1.2 Initial Conditions core Power Level A minimum initial power level typical of a critical, zero power startup condition maximizes the powet excursion.

Prerrurizer Pressure High initial pressurizer pressure maximizes the peak transient pressute.

5-1 i

Prensuri2gr Level

-High initial precsurizar level minimizes the volume of the steam bubble and therefore maximites che pressure increase'following an insurge, hearter-vessel Averade Teocerature Reactor vessel average temperature is not an imt .; art parameter in this analysis.

RCD Flow RCS flow is not a7 important parameter in this analysin.

Core Dvoass Elgg Core bypass flow is not an.important parameter in this analysis.

t E1Aa1 "2nfrator Leve.1

• Initial steam generator level is not an impolcant parameter in this analysis.

ruel Temnerature Due to the zero power initial condition, the initial fuel temperature equal to T-ave. The fuel-clad gap conductivity _is set-high to maximix-

. heat transfer from the fuel.

Steam Generator Tube Pluacing A bounding high tube plugging- value degrades primary-to-secondary heat transfer. 1

. 5 .1.1. 3 ' Boundary Conditions Non-Conductina Heat Exchancers For initialization purposes, non-conducting beat exchangers are used to remove reactor coolant pump heat since the steam generators ar- nassive at-initialization. These are turned off prior _to the start of .he power excursion.

PCP Ocoration Four reactor coolant pumps are in operation to increase the pressure drop around the-loop, and-to minimize thermal feedback during the power excursion. I Pressuri ur Safety Valves

-The pressurizer safety valves-are modeled with lift, accumulation, and blowdown assumptions to maximize RCS pressure during the transient. l Steam Line Safety Valves Although not important for this' transient, steam line safety valves are modeled'with lift, accumulation, and blowdown assumptions to minimize.

primary-to-secondary heat transfer.

l 5-2 i

5.1.1.4 Control, Protection, and Safeguards System Modeling Fert er Tric The high power range flux (Iow setpoint) trip includes a conservative allowance to account for calibration error, and error due to tod withdrawal effects. The response time ci the high flux trip function is the Technical Specification value.

Pies suri zer Pressure centrol Pressurizer spray and PORVs are inoperable to maximize RCS pressure during the transient.

hfpuri2e Level control Due t o t he short C .'* ion of this transient, heaters, makeup anta letdown are unimportant.

Reim Line PORVs and 2 Meneer Ste=tm Dum steam line PORVs and steam dump to condenser are unirnportant for this transient and are inoperable.

5.1.2 core Cooling Capability Analysis 5.1.2.1 hodalization The core cooling capability analysis, which determines the minirnum DNBR, is analyzed with tree reactor coolant pumps in operation. A two-loop model with one single loop and one triple loop is utilized for this analysis. The standard model (Reference 2, Section 3.2) is used with one significant exception. Since this transient initiates at zero power, and since the duration of the transient is very short, the steam generator secoadary response is not important. Rather than using the standard steam generator secondary nodalization, a single secor.dary -

volume is used. The single volume uses the bubble rise option with the local-conditions heat transfer mcdel applied to the steam generator tube conductors. With this modeling approach the initial condition of nero power can be obtainnd, and the primary-to-secondary heat transfer that occurs following the power excursion can be simulated. No main or auxiliary feedwater or initi&l steam flow is modeled. The pres,urizer modeling includes the use of local cor:ditions heat trannf er option for the vessel conductors.

5.1.2.2 Initial conditions Cnre Power Lecg1 A minimum initial power level typical of a critical, zero pcwer startup condition maximizes the power excursion.

Pressurirer Precrure Mominal pressure is assumed, with the pressure initial condition uncertainty accounted for in the Statistira' Core Lesign Methodology.

5-3

- _ _ _ _ -__ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _- _ - - _ - -. - _ _ _a

l -

fierrurirer Level Low initial pressurizer level minim 1:es the pressure increase following an insurge.

Renct or Vergel Averace Tenrerature The nominal temperature corresponding tc zero power operation is assumed, with the tcmporature initial condition uncertainty accounted for in the Statistical Core Design Methodology, .

PC9 Flow Nominal three pump flow is assumed since low flow is conservative for 1 DNER evaluation, The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology.

Cnie EvDass F]ow The nominal calculated flow is assumed, with the flow uncertainty accounted fer in the Statistical Core Design Methodology.

Steam Generator Level Initial steam generator is not an important parameter in this analysis.

(

Fuel Temneratute Due to the initial zero power condition, the initial fuel temperature is equal to T-avo. The fuel-clad gap conductivity is set high to maximita heat transfer from the fuel.

Steam Generator %be Plucainc No tube plugging is assumed to maximize the ECS volume and thereby minimize the incurge into the pressurizer.

5.1.2.3 Boundary Conditions Uon-Conductina Heat Evchancers For initialization purposes, non-conducting heat exchangers are used to remove reactor coolant pump heat since the steam generators are passive at initialization. These are turned off prior to the start of the power excuraion.

ECP operation Since low flow is conservative for DNBR, the minimum nucher of teactor a coclant pumps (three) 1aquired for the modes for which this trr aient is applicable (Modes 2 and 3) are assumed to be in operation.

Pressurirer Safety Valver The pressurizer safety valves are modeled with lift, accumulaticr., and blowdown assumptions to minimize RCS pressure during the transient, steam Lino raferv valvg Although not important for this transient, steam line safety valves are modeled with lift, accumulation, and blowdown assumptions to maximize primary-to-secondary heat transfer.

5-4

3 .-

5.1.2.4 Control, Protection, and Safeguards System Modeling React or Trir, I

The high power range flux (low setpoint) trip includes a conservative allowance to account for calibration error, and error due to rod withdrawal effects. The response time ei the high tiux trip function is the Technical Specification value.

Pressurizer Pressuto contr d Pressurizer cpray and PORVs are operable to minimize RCS pressure during the transient. Heaters are not energized during the transient.

Stem Line PORVs and condenser Ste r Durn Steam line PORVs and steam dump to condenser are unimportant for this transient and are inoperable.

5.1.2.5 other Assumptions Due t o the potential for bottom-peaked power distributions during this transient, and due to the non-applicability of the Statistical Core Design Methodology below the mixing vane grids in the current fuel assembly designs, acceptable DNBRs are confirmed with the W-35 CHF correlation as necessary. Explicit accounting for uncertainties (i.e.

non-SCD) are used with the W-3S correlation.

5.2 Uncontrolled Pnnk Withdrawal at Powei The uncontrolled bank withdrawal at power accident is characterized by an increase in core power level that ceinot be matched by the secondary heat sink. The resultant mismatch causea an increase in primary and secondary system temperatures and pressures. The increases in power and -

temperature, along with a change in the core power distribution, present a DNBR concern. The primary and secondary overpressure limits of 110%

of design pressure are also of concern.

Peak RCS pressure. peak Main Steam System pressure and cote cooling capability are each analyzed separately due to the differences in assumptions required for a conservative analysis. The core cooling capability analysis demonstrates that fuel cladding integrity is main-tained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology.

5.2.1 Feak RCS Pressure Analysis 5.2.1.1 Nodalization Since the transient response of the uncontrolled bank withdrawal event is the same for all loops, the single-loop model described in Section 5~

_ _ _ _ _ - - - - - - - _ _ _ - - - - - _ - - - - - - - - - - - -. .o

2

. 1

~

3.2 of Reference'2 is utilized for this analysis. The pressurizer modeling includes the'use of the local conditions heat transfer option

-for the vessel conductors.

5.2.1.2 Initial Conditions

. Core Power Level' Initial pressurizer pressure and, thus, initial margin to the overpressurization limit is independent of initial power level. Due to ,j the pressure overshoot during the reactor trip instrumentation delay, {

naximum pressure is-achieved with the' maximum prerscrization rate. The maximum pressurization rate is achieved with the maximum insertion of ';

reactivity, provided that ~ reactor trip on high flux does not occur prior to significant system heatup. Since the initial margin to the high flux reacter-trip is greatest at a low poser level, this power level yields the most rapid insertion of reactivity with significant system heatup, Pressurirer Pyessure Initial pressurizer pressure is the nominal value, and the uncertainty in pressure is accounted for in the high pressure reactor trip setpcint. f Pressurizek Level High. initial level'mit.imizes the initial volume of the pressurizer steam space, Jwhich maximizes' the transient primary pressure response.

' Reactor Vescel Averace Temperature Initial ter.perature is not an important parameter in this analysis.

RCS-Flow Initial RCS flowrate is not.an important parameter in this analysis.

Care Evoass' Flow Core bypass flov is not aniimportant parameter in this analysis. '

steam cenerator level High initial. level minimizes the. initial volume of the steam generator steam space, which maximizes the transient secondary pressure response.

Maximum secondary. pressurization causes maximum secondary temperature

' response,-- which minimizes primary-to-secondary heat transf er.

Fuel'Temneraturei Low fuel temperature,-' associated with.high gap conductivity . maximizes

.the transient heat transfer from the fuel to the coolant.

rteam cenerater Tube Plucaina A bounding-high tube plugging value degrades primary-to-secondary heat transfer.

j 5-6 1 l

i i

5.2.1.3 Boundary Conditions s

Pierrurizer raferv Valvos The pressurizer safety valves are modeled with litt, accumulation, and

[ blowdown assumptions which maximize the pressurizer pressure.

l h g Lino Saferv Valyn The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure l

and minimize transient primary-to-secondary heat transfer.

I 5.2.1.4 control, Protection, and Safeguards System Modeling Roartor Trip _

The pertinent reactor trip functions are the overtemperature AT (OTAT),

overpower AT (OPAT), pressurizer high pressure and power range high flux (high setpoint).

The response time of each of the two AT trip functions is the Technical Specification value. The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2 of RefereLee 2. In addition, the AT coeffi-cients used in the analysis account for instrument uncertainties.

The response time of the pressurizer high pressure trip function is the Technical Specification value. The pressuri::er high pressure reactor tilp setpoint is the Technical Specification value plus an allowance which bounds the instrument uncertainty.

The response time of the power range high flux trip function is the Technical-Specification value. The power range high flux trip high setpoint is the Technical Specification value plus an allowance which -

bounds the instrument uncertainty. The high flux signal is adjusted to account for the effects of bank withdrawal.

Presruriner Pressure control In order to maximize primary system pressure, no credit is taken for pressurizer spray or PORV operation.

Pres 9urizer Level control Pressurizer level control system operation has negligible impact on the results of this analysis.

6 Steam Line POFVs and Condenger Steam Dumo Secondary steam relief via the steam line PORVs and the ccadenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.

Steam Generator Level Centrol Feedwater control is in automatic to prevent steam generator low-low level reactor trip.

5-7

- - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~

L TurLirw control ,

The turbine is modeled in the load control mode, which is described in section 3.2.5.1 of Reference 2.

kg M L n _podwater Auxiliary feedwater is disabled. The addition of subcooled auxiliary feedwater would tend to subcool the water in the steam generator, and provide better heat removal capability.

Turbino Trin l

Tutbine tr ip upon r eactor trip is modeled in order to minimize t he post- j trip primary-to-secondary heat transfer, u

=

L 5.2.2 Core Cooling capability Analysis I 5.2.2.1 Since the transient response of the uncontrolled bank withdrawal event tiodalization is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis. The pressuriner modeling includes the use of the local conditions heat transfer option for the ves,el conductors.

5.2.2.2 Initial Conditions core Power Level The uncontrolled bank withdrawal event is analyzed with a spectrum of initial power levels which range from low power to full power.

Uncertainties in initial power level are accounted tor in the Statistical Core Design Methodology.

Pressurizer Precsure Initial pressurizer pressure is the nominal value, and the uncertainty in pressure is accounted for in the Statistical Core Design Methodology.

Pressurizer Level Initial pressurizer level is the nominal value which corresponds to the initial power level, and uncertainties are accounted for in the initial value. Low initial level maximites the initial volume of the pressurizer steam space, which minimites the transient print.ry pressure response.

Reacter Versel Averaae Temerature The nominal temperature corresponding to the initial power level is assu tad, with the temperature initial condition uncertainty accounted for in the Statistical Core Design Methodology.

5-8

RCS Flow The Technical Specification minimum measured flow for power operation is assumed since low flow is conservative for DNBR evaluation. The flow initial co-11 tion uncertainty is accounted for in the Statistical Core Design Me ,odology.

Core F/Dasr Flow The nominal calculated flow is assumed, with the flow uncertainty accounted for in the Statistical Cole Design Methodology.

Steam Generator Level Initial steam generator level is the nominal value which corresponds to the initial power level, and uncertainties are accounted for in the initial value. High initial level minimizes the initial volume of the steam generator steam space, which maximizes the transient secondary -

pressure response. Maximum secondary pressurization causes maximum secondary temperature response, which minimizes primary-to-secondary heat transfer.

Fuel Tenneraturg Initial fuel temperature is the va'.ue which corresponds to the initial power level. Low fuel temperature maximizes the transient heat transfer from the fuel to the coolant.

Sto w Cenorator Tube Plucaina A bounding high tube plugging value degrades primary-to-secondary heat transfer.

5.2.2.3 Boundary Conditions Pressurizer Safety Valves The pressurizer safety valves are modeled with lift, accumulation, and blowdown assumptions which minimize the pressurizer pressure. -

Stoan Line Safetv Valves The steam line safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize transient secondary side pressure and minimize transient primary-to-secondary heat transfer.

5.2.2.4 Control, Protection, and Safeguards System Modeling React or Trin The pertinent reactor trip functions are the overtemperature AT (CTAT),

overpower AT (OPAT), pressurizer high pressure and power range high flux (high setpoint).

The response time of each of the two AT trip functions is the Technical Specification value. The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2 of Reference 2. In addition, the AT coeffi-cients used in the analysis account for instrument uncertainties.

5-9

__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - - . - _ _ . . _ - - _ - _ _ _ _ u

s The response time of the pressurizer high pressure trip function is the

?/echnical Specification value. The pressurizer high pressure reactor '

trip setpoint is the Technical rpecification value plus an allowance which bounds the instrument unceitdr#y.

The response time of the power range high flux trip function is the Technical specification value. The power range high flux trip high f

h setroint is the Technical Specification value plus an allowance which bounds the instrument uncertainty. The higt' flux signal is ad]usted to account for the effects of ba.J withdrawal.

Piercurizer Prennuro Contr:1 l A sensitivity study is performed on pressurizer pressure control. Two I

[

modes are analyzed, one in which pressurizer pressure control is in manual with sprays and PORVs disabled, and the other in which pressur-izer pressure control is in automatic with sprays and PORVs enabled. .

I Pierrurizer Tevel Control Pressurizer level control is in manual. Level control has negligible impact on the results of this analysis.

Eream Line PORVs and condenser Ste r Dumo Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in crder to maximize secondary side pressur zation and minimize transient primary-to-secondary heat transfer.

Stear Generator Level Contrcl Feedwater control is in automatic to prevent steam s enerator low-low level reactor trip.

Turbine Control The turbine is modeled in the load control mode, which is described in Section 3.2.5.1 of Reference 2.

Auxiliary Feedwater Auxiliary feedwater is disabled. The addition of subccoled auxiliary feedwater would tend to subcool the water in the steam generator, and provide better heat removal capability.

Turbine Trin Turbine trip upon reactor trip is modeled in order to minimize the post-trip primary-to-secondary heat transfer.

5.3 control Rod Misoreration I st at ic al h Misaliannd Pod)

The statically misaligned rod event considers the situation where a control rod is misaligned from the remainder of its bank. A rod misalignment may produce an increase in core peaking which decreases the margin to DNS. Steady-state three-dimensional power peaking analyses are performed to confirm that the asymmetric power distributions resulting from the rod misalignment will not result in DNB. There is no system transient associated with the analysis of the statically 5-10

misaligned rod case. The reactor is assumed to remain at its initial power level.

The statically misaligned rod evaluation is performed at nominal hot full power (HFP) conditions. Axial shapes allowed by the power L dependent AFD limits cre considered in the evaluation. Two specific cases are analyzed which characterine the worst case misalignment s . The first case considers the full insertion of any ona rod with Control Bank D positioned anywhere within the full power rod i ,errion limits (RILs).

J.. s'"'snd c&se considers the misalignment of a single Control Lank D rod at its fully withdrawn position, with the remainder of Control Bank

( positicaed at the full power rod insertion limit.

[

Power distributions resulting from Case 1 are not analyzed for each reload core. This is because the thermal conditions (reactor power, -

pressure and coolant temperature) rnd power distributions evalu'ted in the dropped rod transient analysis bound the thermal conditiens and power distributions that would occur in the statically misa11gned rod event describe- 49 Case 1. The asymmetric power distributions resulting from Case 2 are ve "ated for each reload core to ensure that the minimum DNBR remains above the 95/95 DNER limit based on acceptable correlations. The minimum DNBR is determined using the Statistical Core Design Methodology.

The peak linear heat generation rate produced from the rod misalignment is confirmed for each reload core to be less than the linear heat generation rate which would result in fuel melt. The peak linear heat generation rates resulting from rod misalignments do not challenge the fuel melt limit.

5.4 Control Pod Mircreration (?inM e Rod Withdrawal)

The single rod withdrawal accident is characterized by an increase in the power generation of the primary system, and since the heat removal capability of the secondary system is not increased during the tran-sient, the resultant power mismatch causes an increase in primary and secondary system temperature and pressure.

The acceptance criterien for this event is to ensure that there is adequate core cocling capability. The coro cooling capability analysis determines to what extent fuel claddina integrity is compromised by calculating the number of fuel rods that exceed the 95/95 DNBR limit based on acceptable correlations.

5.4.1 Nodalization Since the transient response of the single rod withdrawal event is the same for all loops, the single-loop model described in Section 3.2 of Reference 2 is utilized for this analysis. The pressurizer modeling includes the uso of the local conditions heat transfer option for the vessel conductors.

5-11

____ - __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ ___ - _ _ -a

4 r

. 5.4.2- -Initial Conditions

' core Power Level-Initial power is the nominal full power value. Uncertainty in power level is accounted for in thefstatistical Core Design Methodology.

Pressurizer Pressure

- Initial-pressurizer pressure is the nominal value. Uncertainty in pressure is-accounted for inithe Statistical Core Design Methodology.

Pressuricer L'evel

. High initial level minimizes the initial volume of the pressurizer steam-

]

space, which maximizes the transient primary pressure response. Up to the limit of the ability of the_ pressurizer sprays to control pressure, 3 maximum pressure-js' conservative in order to delay reactor trip on OTAT.

Reactor Vessel Averaae Temnerature Initial temperature is the full power nominal value. UncertaintyLin this parameter is accounted _for in the Statistical Core Design Methodology.

Rcs-Flow The Technical Specification minimum measured flow for power operation is-assumed since low flow is conservative for DNBR evaluation. The flow initial condition uncertalnty is accounted for in the Statistical Core DeLign Methodology..

Core Dynass-Plow The nominal calculeted flow is-assumed, with the flow uncertainty

~ accounted for in the Statistical Core Design Methodology.

Steam Generator Level High initial level minimizes the' initial volume of the steam generator

- steam. spaces which maximizes-the transient secondary pressure response.

Maximum secondary pressurization causes' maximum secondary. temperature

- response, which minimizes primary-to-secondary heat transfer-,

- Fuel Temnerature Low fuel temperature, associated with high gap conductivity, maximizes the transient heat: transfer from the fuel to the coolant.

Steam Generator Tube Pludoinc.

- A bounding high tube- plugging value degrades primary-to-secondary heat 1

' transfer.

l-5.4.3 Boundary Conditions-Pressurizer Saferv Valves j The pressurizer safety valves are modeled with lift, accumulation, and -j blowdown assumptions which minimize the pressurizer pressure.  ;

J 5-12

rt eam Line sa f et y valves s The steam line safety valves are modeled with lift, arcumulation, and blowdown assumptions which maximize transient secondary side pressure and minimize transient primary-to-secondary heat transfer.

5.4.4 Control, Protection, and Safeguards System Modeling Egaetor Trin The pertinent reactor trip functions are the overtemperature AT (OTAT),

over power AT (OPAT), pressurizer high pressure and power range high flux (high setpoint).

Tne response time of each of the two AT trip functions is the Technical -

Specification value. The setpoint values of the AT trip functions are continuously computed from system parameters using the modeling described in Section 3.2 of Reference 2. In addition, the AT coeffi-cients used in the analysis account for instrument uncertainties.

The response time of the pressurizer high pressure trip function is the Technical Specification value. The pressurizer high pressure reactor trip setpoint is the Tecnnical Specification value plus an allowance which bounds the instrument uncertainty.

The response *1me of the power range high flux trip function is the Technical Specification value. The power range high flux trip high set poir.t is the Technical Specification value plus an allowance which bounds the instrument uncertainty. The high flux signal is adjusted to account for the effects of rod withdrawal.

Pressurizer Pressure Centrol Pressurizer pressure control is 'n automatic with sprays enabled and PORVs disabled in order to delay teactor trip on OTAT and delay reactor trip on high pressurizer pressure.

Precrurizer Level ConM Pressurizer level control is in manual with the pressurizer heaters disabled in order to delay reactor trip on high pressuriner pressure.

Charging / letdown has negligible impact.

St eam Line POPVs and Condenser Steam Durn Secondary steam relief via the steam line PORVs and the condenser steam dump is unavailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary heat transfer.

Steam Generater Level Centrol Feedwater control is in automatic to prevent steam generator low-low level reactor trip.

5-13

1 Auxilials leM. A ';

s Auxiliary fo9 h ies is disableJ. The addition of subcooled auxillary l t eedwat er would t ond t o sulmool t he wat er in the steam generator, and '

reduce secondsty s.ide prersure.

I l

Tu! L ing2M J The t ut bino is moduled in the load cont r ol rode, which is described in Section 3.2.5.1 of Peterence 2. s Tutbine 71io Turbine trip upon reactor tiip is inodeled in order to minimize the pcrt -

t rip pr itury-to-secondar y heat transler.

5.5 rinrtue of An Inrt ive Fort er coolant Pi!" n At 7n Inzerrect Z{goerature  :

i The itGuite and Catawba plant Technical Specificatiors cerrentl',' require that all f our RCPs be t unning at power operat ion. Fur t he r rnol e , low flow

• i any RCS . cop, coincident with reactor power above the P-d interlock

(:uttently at 48% of rat ed thermal power) will caust o reactor trip.

Therefore, the only situat ion in which the subject icident is possible l 1s a ttip of orie RCP below P-8. For this sitts on ...e operator might l choose, during allowable at power outage time for the fourth RCP, to at t empt a restart of the tripped pump. The accident is analyzed from ,

the most conservative condirion allowed by the Reactor Protection l Dystem, even though operator error is required for the analyzed I scenario to occur. The acceptance criterion is that fuel cladding integrity shall be maintained by ensuring that the minimum DNER remains the above the 95/95 DNBR lit..it based on acceptable correlations.

5 . c. 1 Nodalization Decause at the loop asymmetry between the inactive sinale loop and the three active loops, the double-loop RCS model described in Section 3.2 of Reference 2 is used. The pressurizer modeling includes the use of the local conditions heat transfer option for the vesse conductors.

5.5.2 Initial Conditions Qre power Since the positive reactivity insertion due to the colder moderator average temperature causer a power increase, the initial indicated power level must be suff.iciently less than the P-8 setpoint such that, by the time ro s. indicated power level reaches the P-8 setpoint, the indicated flow in tie loop containing the restarted pump is greater than the low flow reactt,t trip setpoint. This delays or preventr reactor trip and is theref ore conservative f or DNBR evalt.ation. l frersuri;er Presrure A >ressure initial condition uncertainty is applied to minimize pressure dut!ng the transient since this .is conservative ;or DNER evaluation.

5-14

pierrurirei 1_t ve.

The heatup of the colder water and the increase in core power will cauro an expansion of the teactor coolant art an increase in prescuritet a level. Is negat ive level uncert ainty is used in ordet to naximize the size of the piescurl:er steam bulble to be c o:tp r es s ed , which rainimizes the tzansient prescute recponse.

Met ot vu r e1 Ic/e r rae 'I er r u aL=

A positivo t ouperature uncer t ainty is used t o minimize the mar gin to D!ilt .

EN Flow in order to minimize core flor, arid theref ot e the nargin t o D!Jb, the thtoo pump equivalent of the Technical Specification minimum measured flow is adjunted by a negative flow uncertainty cro I"co ar s Flev A positive tiow uncertainty is used to ininimite the marcin to Ll!B.

Steam Generater Lo"ol The results of this transient are not sensitive to the direction of steam generator level uncertaint: as long the transient level response is kept within the tange that avoids protection or safeguards actuatien.

Fucl Temneraturq A low initial temperature is assumed rnaximize the gap conductivity calculated for steady-state condit ions and used f or the subsequent t r ansient . A high gap conductivity minimites the fuel heatup and at t enda t negative reactivity in,ertion caused by the powet increase.

This makes t he powet increase more seveze and is therrfore conservative f or DtJB evaluation.

Eigeam Generator Tube Pluanina -

Steam generatot tube plugging is not an important parameter in this analysis.

5.5.3 Loundary conditions EcP orseration The RCPs operating prior to the accident are modeled ossuming constant speed operation thtoughout the transient. The Kir' that 18 inactive at the start of the accident is modeled with a conservative speed vs. time controller.

5.5.4 Control, Protection, and SafeguatJs Systems Modeling Prerturizer Precure control The prescuriter sprays and POR\'s are assumed to be operable to minimize the pressure increase resulting from the pump restart and power increase.

5-15

., .__m ___.-m__._... _ _ _ _ _ . . _ _ _ _ ___ _ _ _. _ _ m_ -

Preneurizer Level control Ho credit is taken for pressurizer hee.ter operation to compensate for the increase above programmed pressurizer lovel which occurs due to the power increase. Heater operation would tend to elevate pressure.

Ete1m Generator Lovel control Automatic level cont rol is assumed to allow the reactor to reach a higher power level while minimizing the possibility of tripping on low-low steam generator narrow range level due to increased heat input irom the primary system.

Inv l'en. reced cent r el Aut omatic - speed control is assumed to avoid tripping on low-low steam generator narrow range level due to increased-heat input Irom the primary cyctem. ]

Fod Cent M The Rod control System is assumed to be in automatic when reactor vessel {

average temperature decreases. The temperature dierease will cause rod i withdrawal and an increase in core power .

Turbino control The turbine is assumed to be in manual control. In this mode, the valves do not respond to changes in steam'line pressure. Therefore, when steam line pressure increases due to increased heat input from the primary system,.the steam flow to the turbine will increase. This-will retard the core power less than if the turbine control valves closed down and caused steam line pressure and RCS temperatures to increase further.

Auxiliarv Feedwater AFW flow would be credited when the safety analysis value of the low-low ,

- steam generator level setpoint'is reached. However, the parameter of interest for this transient has reached its limiting value before the appropriate Technical Specification response time delay'has elapsed.

- Therefore, no AFW is actually delivered to the steam generators.

l 5.6 CVcr Malfunction That Results In A Decreare In Poron concen-tration In The Reactor Coolant A baron dilution occurs when the-soluble boric acid concentration of makeup _ water supplied to the RCS is less than the concentration of the existing reactor coolant. The boron dilution accident postulates that

. ruch a-dilution occurs _without adequate administrative control such that ,

there was the potential for loss 4 shutdown margin. This accident is '

conservatively analyzed to ensuts t the diNtion is terminaced, . by manual.or automatit means, within a.,ropriate time limits. In accordance with Reference 3, appropriate time is judged to be at least 15 minutes for: Modes 3-5 and at least 30 minutes for Mode 6.  !

The licensing bases for the McGuire and Catawba Nuclear. Stations are different. .. For McGuire, this accident is analyzed for the power '

operation (Mode 1), startup (Mode 2) ,- ad - refueling (Mode 6) rodes of i ,

i l

l 5-16

operation. Ihnual oper at i,;n i t relied on to t erminato the dilut ion in all t hr ee modes . For Cat awba, this accident is analyzed for the power t operation, stattup, hot standby M8 ode 3), hot shutdown (Mode 4), cold shutdown (Made 5), and r ef uelira raades of cperatien. Aut omat i c creration of the bot on Dilution Mitigation System (bDMS) is ielied on to t er minat e the dilutjon in hot s t, o nd b/ , hot shutdown, cold shutdown, and ref ueling, with rnanual orm d en _ substitute treans when the DL!E 2n inopetable. Manual 0peration ir t eiled on to terminate the dilution in powet operation or st at tup.

The various modes at the two stat ions are analyzed wit h two dif f er ent methods for two different purposes. S'i t e t , with the BDMS applicable and asrumed to be operable, the accident is analyzed to demonst rate that there 10 adequat e t ime, without restrictions on t he f l ow l at e s from potent ial dilution sources, for the BDMS to teindnate the dilution prior __

to criticality. This time consists of two comporents: 1) the period required to stroke the valves nanipulated by the DDMS and 2) the period i requited, once the unborated water cource has been isolated, to putge the temaining unborated water from the piping leading to the RCS.

Second, with the PDMS inapplicable or assumed to be inoperable, the accident is analytod to demonstrate that there is adequate t irne, possibly with restrictions on the flow rates from potential dilution sources, f or t he operator to terminate the dilution prior to criticality. Since the EDMS is not used in Modes 1 and 2, the analycit of these modes is similar to the analysis o: Modes 3-6 with the LDMS assume,1 to be inoperablo, but without the rect rict ions on f low l at es During Mode 6 an inadvertent dilution Irom the Reactor Makeup Water Syntem is prevented by administrative controls which isolate the RCS from potential sources of unborated makeup water. The results of the accident analysis for this mode ate for an assumed dilution event, for thich no mechanism or flow path has been identified. The results of the accident analysis are for the dilution flow rates which, assuming the boron concentrations are at the teload safety analysis limits, give -

exactly the acceptance riteria operator response times. Flow rates are testricted, through Technical Specifications and administrative controls, to valuen which are less than these analyzed flow rates, thus in pract. ice giving even longer operator response times. Additional mat gin j a provided by the fact there is typically margin between the assumed boton cencentrations for a given mode and the actual corresponding concentrations for the roload core.

5.6.1 Initial Conditions Dilution Volume A postulat.ed dilution event progresses faster for smaller KCS water volumes. Thcrefore, the analysis considers the smallest RCS water volume in which the unborated water is actively mixed by forced circulation. For Modes 1-3, the Technical Specifications require that at least one reactor coolant pump be operating. This force? circulation will mix the RCS inventory in the reactor vessel and each of the four reactor coolant loops. The pressurizer and the pressurizer surge line are not included in the volume available for dilution in Modes 1-3. For 5-17

r normal operation in Mode 4, f ot ted circulation is typically naintained, alt hough the Technical Specif ications do not require it. The volume available for dilution in Mode 4 is therefore conservatively assumed to not include the upper head of the reactor vessel, a region which has reduced flow in the absence of forced circulation, or the pressurifer and the pressurizer surge line. Since the Technical Specifications do require operability of all four steam generators during Mode 4, all fout of the reactor coolant loops, in addit ion to the remainder of the -

t eactor vessel, are included in the Rcs volume available for dilution.

For Modes 5 and 6, the reactor coolant water level nay be drained to below the top of the main coolant loop piping, and at least one train of the Residual Heat Removal Syctem (RHRS) is operating. The volume available for dilution in these modes is limited to the snallet volume RHRS train plus the portions of the reactor vessel and reactor coolant loop piping below the minimum water level and between the RHRS inlet and outlet connections. The minimum water level used to calculate this volume is correctod-for-level inottument uncertainty.

Doron cencentiatione The Technical Specifications require that the shutdown matgin in the variouc modes be above a certain minimum value. The difference in boron concentration, between the value at which the relevant alarm function is

-actuated and the value at which the reactor is just critical, determines the time available to mitigate a dilution event. Mathematically. this timo is a function of the ratio of these two concentrations, where a large ratio corresponds to a longer time. During the reload safety

-analysis for each new core, the above. concentrations are checked to ensure that the value of this latio for each mode is larger than the corresponding ratio assumed in the accident analysis, Each mode of operation covers a range of temperatures. Therefore, within that mode, the temperature which minimizes this ratio is used for comparison with the accident analysis ratio. For accident initial conditions in which the control rods are withdrawn, it is conservatively assumed, in calculating the critical boron concentration, that the most reactive rod

_does not fall into the core at ieactor trip. This assumption is also conservatively applied in Mode'3 when the initial condition is hot zero power. For colder conditions in Modes 3-5, emergency procedures for reactor trip with a stuck rod require that, prior to the initiation of 4 the cooldown, the coron concentration be increased by an amount which compensates for any rods not completely inserted.

l 5.6.2 Boundary Conditions In the absence ot flow rate restrictions, the dilution flow rate assumed j to enter the RCS is greater than or equal to the design volumetric flow i

-rate of both reactor makeup water pumps. In a dilution event, these pumps'are assumed to deliver unborated water to the suction of the centrifugal charging pumps, :Since the water delivered by these pumps is typically colder than the RCS inventory, the unborated water expands within the RCS, causing a given volumetric flow rate measured at the colder temperature to correspond to a larger volumetric dilution flow rate within the RCS. This density difference in the dilution flow rate in accounted for in the analysis. The above assumption on flow rate is 5-18

}

aloo consi i vat Ively used ict MTh 6. Any nakeup which 1E 1 kuit ed du Ang tht: u de it tot at t d wat er nu;. ;.l u d f r on, t he 1e:w litg w.ttet storage tank.

4.6.3 cent i el , Pr ot ect ion, and ?ateguat h rycten! Mob 1tng Mit igat ion of a boton dilution acu dent it nct assurred t o begin until an alarm han w u ned of the abnot aal :it run.st ancet caused by tae event. Far M:vinc 3-6 wit h t he I D!C o[+ at:le , the alain f unct ion i.I provided l'y the moasured tource range count rate excr eding the PDM" set roint .

Fct !bde:

3-6 with t he toms inoperable, t he alarm tunction is pr ovided by ' he source t attge high- flux-at -shut down olat a ext eeding it s betroint For M>de 2 and for manual tod control duting Mode 1, the alarm function is 1stovided by the eat 1 test Ieactor t r ip cet point r e-a c h ed . Fjnally, ior -

aut emat ic rod cont Iol dur ing Mode 1, the alain functicn is ptovided by the alarm which occurs when the control t oda reach their inncition limitc.

s.7 1ru do r t e nt Lcadina ed Opei nt ien a f Flel Ic N_ n ! l v I n 7gl In ct er e r Ioriticn rote loading errots can occur from the improper loading of one or more fuel assemblien in an improper posit ion, from enrichu nt errors, or from the misloading or omission of burnable absolbor tode. The result of these errors is the possibility that core peaking will exceed t he pe'tkirig calculat ed f or the correct cote loading.

Administ r ative pt oceduren at e in place t o prevent enrichment errors during fuel fabrication and during cote loading. Also, a rigotous startup physics teuttng program is performed r.ubsequent to each core loading that would detect any credible misloaded fuel assently The i micloaded fuel assembly analysis confirms that the increase in peaking ptoduced frcm a loading error or enrichment errot would either be detected by the incor e flux mapping system, or would be less than the peaking uncertainties included in the analysic of both Condition I and Condition II events.

5-19 a

6.0 l u :EF A:;O IN ELACTSF CZLANT l!TJL!; TORY 6,1 Ir.2dve r t e nt mm.:,.n' t ' ' ' " " ' - t' c 'q s i aticn The irwivet t , nt cj.a ation of the lh 2 P. nr/ Cct e t c ollna : ysti: ould be ca c od Lc/ eit het treratot et r ot ci a m il l ou n electrical a t uat icn cignal. U;; n r eceipt of t he actuat2tn si'inal, the cent l i t ug al char ging pun.; c l egin d-liva log highly bor at ed r etueling water st orage tank water ta the Peactet Coolant Cyctem 'Jh e : asultant negative Ieactivity inGettton Cause. d deC t ea f-e An Cole pC4et and, CCnDequently, d deCIea9e in t empe r at u t e . Coolant shr ink age causec a reduct. ion in bot h presnurizer watet level and pra t ure IJJb is the prinar y concet h tot thin trannient due to this decrease in syctem pressure.

The inagn i t ude of the pressure decrease for this transient is no mat e nevete than that f or t he inadvet t en; opening of a ptent.urizer safety et relief valve t r ansient , which also t r ipu the reactor cn low pressurizer pressure. Fu r t hel n. ore , the opening of a safety valve does not :ntroduce the core power and Reactor Coolant Syst em t eapel at ure decreates that are char act et ist ic of the inadvertent ECC.? actuatlon. INither event involves any r educt ion in the heactor Coolant Syst em f lowrat e , since the leactol coolant punp: ate not tripped. Therefore, the Df;b resultn of this trancient are bounded by the inadvertent opening of a ptessuritet saf et y or Iel.ief valve transient.

Laued on t he above qualit at ive evaluat ion, a quantitative analysis of thin transient ir not required. Chauld a teanalysis b4_ccme necessaty, eithet due to plant changes, modeling changes, or other changes which invalidate any of the above argun1-nts, the analyt ical nathodology enployed would be as follows.

The cor e cooling capability analysis demonstrates t l .a t fuel cladding int egr ity is uintained by ensuring t hat the minitnum DNbi< remains above -

the 95/95 DNLR limit L>ased on acceptable coIrelations. The minimum O!aR in det etmined using the Statistical Core Design Methodology.

ts.1.1  !!odali zat ion Since the inadvertent ECCS operation translent is cyrsetrical with I e nac t to t he f our react or coolant loops, a cingle-loop model (Reference 2, recticn 3.2) is utilized for thic analysis. The pleocutinet modeling inclucec the use of the local conditions heat tr ans f er ept ion f or t he vessel conductore.

6.1.2 Initial Conditionc Qi e it.:er Level High init ial power level maximizes the p1imary systeri heat ' flux. The uncertainty in this parameter is account ed tot in the S t a' tical Core Design Methodalogy 6-1

Prerrufizer Precrurg Nominal full power pressurizer pressure is assumed. The uncertainty in this parameter is accounted for in the statistical Core Design ]

Methodology.

Precrurizer Lecel 111gh initial level minimizes the volume of the pressurizer stemn space which maximizes the pressure decr eaue resulting f rom the outsurge.

Reactor vercel Averaae Te merature Dominal full power vessel average temperature is assumed. The uncertainty in this parameter.is accounted for in the Statistical Core Design Methodology.

RCC Plow i The Technical Specification minimum measured flow f or power operat. ion is assumed since low flow is conservative for DUBR evaluation. The flow initial condition uncertainty is accounted for in the Statistical Core Design Methodology, core EYDarc Flow The nominal calculated flow is assumed, with the flow uncertainty f account ed f or in the Statistical Core Design Methodology.

steam cenerator Level Steam generator level is not an important parameter in this analysis. I Funi Tenrier a tu r e A high initial temperature is assumed to minimize the gap conductivity calculated for steady-state conditions-and used for the subsequent transient. A low gap conductivity' minimizes the transient change in  !

fuel rod surface heat flux associated with a power decrease. This makes the power decrease less severe and is therefore conservative for DNBR evaluation.

Steam Generat or Tube Plutmino Steam generator tube plugging is not an important parameter in this analysis.

6.1.3 Boundary Conditions i ECOS Plow A maximun safety injection (lowrate along with a conservatively high  ;

boron concentration yields the most limiting transient response. In ]

. order to_ minimize the delay in'the-delivery _of the borated injection J water, no credit.is.taken for the purgo volume of unborated water in the injection lines'.

Tream Line Safety Valves The-main steam code safety-valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer, 6-2

4 6.1.4 Control, Pi ct ect iot,, and cafeguards Syst on Modeling 9

Pe rt er Trip J%ac t o: trip is dEbumed t o occut on low pressutizer pressure, att er an al tt or t iat e ins t r ument at ion delay.

li m "ut i rei 6 : rule control Pt essui1:or sprays an:1 PORVs are assurwd to be opetable in order to rninimi ze the system prescute throughout the tra%2ent.

h esnur ire r Levol centrol PlessuitZer heaters ata accurned t o be inoperable so that React or Coolant System pressute is minimized. Charging / letdown has negligible impact.

aeu Line POP h and Condenrer "t e rn Dum Secondary steam relief via the steam line PORVs and t he condenser steam durnp is un3vailable in order to maximize secondary side pressurization and minimize transient primary-to-secondary he transfer.

rt eam cener2 Lor Level Centro _1 The r esults of t his transient are not sensit ive to the mode of steam generat or level cont rol as long the level is kept within the t ange thst avoids protect ion or saf eguards actuation.

MFM Pumr, Ctsoed Cent rol The results of this t r ansient are not sensitive to the mode of MFW purip speed control as long as the st eam generator level is kept within the range that avoids protection or safeguards actuation.

Itod cent rnl 11 0 credit is taken for the operation of the Rod Control System for this trannient, which results in an decrease in RCS temperature. Wit h the Rod Control System in automatic, the control rods would cause a positive -

reactivity addition as they are withdrawn in an attempt to maintain RCS temperaturo at its nominal value. The resultant power increase would retard the system depressurization. 5 Turigne control The turbine is modeled in the load control modo, which is described in Section 3.2.5.1 of Reference 2.

Auxi1iary F'eedwat er APW flow would be credited when the safety analysic value of the low-low steam generator level setpoint is reached. However, the parameter of interest for this transient has teached its limiting value before the appropriate Technical Specification response time delay has elapsed.

Therefore, no AFW is actually delivered to the steam generators.

Tutbine Trip The teactor trip leads to a subsequent turbine trip.

6-3

.O

7.0 L E C l;EA.' E Ill PEACTCE CCCUJ!T I!TJLI:T3Y 7.1 I toff. t t c nt

't enirn n_ Piorruiirei rif e E r_u : M ' " ' '

The lon, of invent ory t hr ough the c;'en v21ve caums a deptessurl ation of the PC: , linre the cete ;: owe t , f1ow, and temperatute are ielatavely unalf oct ed pr ior to 1eactot tiir hy this depressuiization, the ieduction in 1,1ersute causer a i eclact ion In ENB rat gin. The al plical>1e acc.-} t anc e cr iter ion is t hat fuel clat-ling integrity shall be maintained by enauting that the n.iniraum LMPR r e:uins the above the 95/90 LNER limit i based on acceptable t trelations. The minimum EM R is dctermined using the JLatistical Cote Decign Methodology.

7.1.1 Moda11:ation Since the valve ol,ening is in the prescuriner, it affects all PCS loopr identically. Ther ef or e a single-loo;; RCS syst em model is used. The pressurizet modeling includes the use of the local conditions hr. a t transfer option for the vessel conductors.

7.1.2 .ut al conditions p er Le el Full powet is assumod in order to maximize the primaty system heat flux.

Tho uncert ainty in this parameter in account ed f or an t he Statistical Cot e Dou lgn Methoctology.

Preccurizer Pierrure

!!aminal pressure ie assumed, with the pressure initial condition uncertainty accounted tot in the Statistical Core Design Methodology, frearutirer Level Since tl.is accident involves a reduction in RCS volume due to inventory locc, a positive level uncettainty is assumed to minimize the initial pressurizer steam bubble volume and therefore m uimine the pressure declease due to invent ory loss.

Mrt or Verrel ' vor re Temerat ur e The nominal temperatute corresponding to full power operation is assumed, with the temperature ini*.ial condition uncertainty accounted for in t he St atist i cal Cole Design Met hodolcs;y.

ECS Finw The Technical specification minimum measured flow for power operation is assumed since low flow is conservative for DNDR evaluation. The flow init ial condition unce. tainty is accounted f or in the Statistical Core Design Methodology, Me Pytu r 9 Floy The nominal calculated flow is assumed, with the flow uncertainty accounted for in the St atistical Core Design Methodology.

7-1

_ -- - . _ _ - _ _ _ - . _ - - - _ - _ . - _ - - _ _ _ _ _ . _ _ . . - - - _ . _ _ ..-- _ _ ___._.-_J

J Etenm Generater Level The 2 esults of this transient are not sensitive to the direction of st eam generator level uncertainty as long the transient level responso is kept within the range that avoids protection or safeguards actuation.

Eyel_ Tenveratute A high initial ternperature is assumed to minimize the gap conductivity )

calculated for steady-state conditions and used for the subsequent l transient. A low gap conductivity minimizes the transient change in fuel tod surface heat flux-associated with a power decrease due to moderator density. This makes the power dc rease less severe and it therefore conservative for DNBR evaluation.

rt eam cenorator Tubo _ D1uadina 4 The results of this analysis ree not sensitive to the amount of steam generator tube plugging.

7.1.3 Boundary Conditions Eteam Line raferv Valvga The steam line safety valves are modeled with setpoint drift, accumulation, and blowdown assumptions which maximize the transient secondary pressure and therefore minimize secondary side heat removal.

7.1.4 control, Protection, and Safeguards Systems Modeling Reactor Trio Reactor trip is on either low pressurizer pressure or overtemperature AT. The Technical Specification response timer are used and the safety analysis setpoints include the effects of uncertainty in the monitored parametar'and in-the setooint.

freenurizer Preocure Control No credit is taken for pressurizer heater operation to compensate for the decream in pressurizer pressure which occurs due to the. inventory loss. This results in a lower post-trip pressurizer pressure, which is conservative for DNER evaluation.

Steam'Cenerator Level Control The results of this transient are not sensitive to the mode of steam generator level centrol as long the level is kept within the rango that avoids protection or safeguards actuation.

MPW Pumn rtseed control- ~

The results_of this_ transient are not sensitive to the mode c- MFW pump speed control.as long as the steam generator level is kept within the range that-avoids pa'>tection or safeguards actuation.

7-2

- -.. . . . . - . - - - - - - -.._--- - . -.. - . - - ~

n rter Gorg_ r ot oi Level Cent rol The r esults of t his transient are not sensitive to the mode of steam generator level control as long the level is keet within the tange that avoids protection er safeguards actuation. a M1.Limr Erm >d "t nt i el -

The r ecult s cf t his t ransient ato not sensitive to tne mode of MFW pun p speed control as long as the st m generator level is kept wit hin t he lange that avoids protection or categualdi, actuation.

Red centiel A penalty is takea for automatic rod control to insert positive t eactivity t o incr ease power and reactor vescel aver age temperature.

These partuneters would otherwise decrease in response to the negative teactivity inserted try t he mode rat or density reduction. _

Tuilsino Conti el Tho tutbine 10 modeled in the load cont rol inode, which is described in Section 3.2.5.1 of Reference 2.

Auxiliaiv Feedwater AFW flow would be cr edited when the saf ety analysis value of the low-low steam generator level setpoint is reached. However, the parameter of int e rest for this transient has reached its limiting value before the appropriate Technical Specificaticn respon.;e time delay has elapsed.

Therefore, na AFW is actually delivered to the steam generat. ors.

Turbino Trip The turbine is tripped on reactor trip. No time delay is assumed since t hic assumption minimizes poot-trip primary-to-secondary heat removal.

7 . .' f t eam Cenerat er '"ubc Rut,ture 2

The steam generator tube rupture analyzed is a double ended guillotine break of a cingle tube. This transient is evaluated in two parts; first to evaluate minimum DNBR, and seccndly to provide offcite dcse input data for a separate evaluation to determine whether the fission product release to the environment is within the established dose acceptarice c11teria.

The DNBR analysis fol this transient is modelod as a complete loss of coolant flow event initiated from an off-normal condition, using tne Statistical Core Design methodology. The loss of flow ia assumed to occur subsequent to the OTAT reactor trip caused by the steam generator tube rupture depressur$ ation.

The initiating event for the offsite dose input analysis is tne double-ended guillotine break of a single steam generator tube. This analysis generates the offsite steam release boundary condition for the doce evaluation. The single failure identified for maximizing offsite dose is the failure of the pORV on the ruptured steam generator to close. In this analysis, this valve remains open until operator action is taken to

'7 - 3

- _ _ _ _ - _ _ _ _ _ _ _ _ _ - - - _ _ _ _ . --___-____ . - _ _ - _. _ . _ - A

I 1solate the PORV. In addition the offsite dcse acc(ptance criteria l

f or thlb t ransient , ol et ator action t o mitigate the trancient and t e t tuinat e primary leakage naast prevent overfillitig t he ruptured st+am ger.et at or .

7.2.1 Cot e tooling Capability Analysis 7.2.1.1 13oda 11:a t i on Since the complete loss of ilow transient is sym::,etrical wit h respect to the four teactor coolant loops, a single loop rrodel (Ecforence 2, Section 3.2) 1s utill:ed for this analysis.

7.2.1.2 Initial Conditionc core Power Level High initial power level maximizes the primary system heat flux. The uncertainty in this parameter is accounted for in the Statistical Core Des!gn Methodology.

Prerrurizer Pi en sur e IJominal pressurizer precsure is a ud. The uncertainty in this parameter is accounted for in the Listical Core Design Methodology.

hygrurizer Level Low initial level increases the volume of the pressurizer steam space which miniinizes the pressure increase resulting f rom the incurge.

Peactor verrel Averane Temperature rJominal vessel average ten.perature is assumed. The uncertainty in this parameter is accounted for in the Statistical Core Design Methodology.

ECO Flow Minimum measured Reactor Coolant System flow is assumed. The uncettainty in this parameter is accounted for in the statistical core Design Methodology.

Corc Dvrars Flow IJominal full power bvinss flow is assumed. The uncertainty in this parameter is accounted ' I in the Statistical Core Design Method 31ogy.

Etenn Generater Leyel Initial steam gener atot level is not an iraportant paralaeter in this analysis.

7-4

Fuel Trt eratur, A high initial t enper at u r e is assu;.ed t o minimize t he gap conduct ivity calculat ed 1c1 steady-state conditione and used for the subsequent transient. A low gap conductivity minimizes the transient change in fuel rod surface heat flux associated with a power decrease. This nakes the power decrease less sev.zre and is therefore conservative fot D!IER evaluatjon.

E t m % n g I a t e r "'td x Pluquira;;

For transients of such short duration, steam generator tube plugging doec not have an etfect on the transient results.

7.2.J.3 Boundary conditionc PCP meration All four teactor coolant pumps are tripped on the loss of offsito power.

The pump model is ad]usted such that the resulting coastdown flow is conservative with respect to the flow coastdown tent data.

Stem Line saf ety Va tzn The main steam code saf ety valves are n'odeled with lif t , accumulation, und blowdown assumptions which maximize secondary side pressure and minimize primary-to-secondary heat transfer.

Offsite Powel; Offsite pcwor is assumed to be lost coincident with ttrbine trip. This isolates steam flow to the condenser, thereby maximizing the atmospheric steam releases.

7.2.1.4 Control, Protection, and Safeguards System Modeling Peacror Trin Reactor trip it assumed to occur on overtemperature .iT, af t er an appropriate instrumentation dela).

Prescurizer Pre;sure centrol Pressurizer sprays and PORVs are assumed to be operable in order to minimize the system pressure throughout the tranrient.

Pressuriner Level cont h Pressurizer heaters are assumed to be inoperable so that Reactor Coolant System pressure is minimized. Charging / letdown has negligible impact.

Steam Line PDRve and condenser et eam The main steam PORVs and condenser dumps valves are assumed to be unavailable during this transient. This maximizes the secondc.ry side pressure and temperature and therefore reduces primary-to-secondary heat transfer.

7-5

St eze Generat or k/e1 Crntiel The results of this tr ansior.t are not sensit ive t o t he n. ode of -team gener at or level control as iong the level is kept within the range t hat avoidc protection or safeguards actuation.

MFM Pur o r t e e d Cr,.p t I o l The results of this transient are nct sensitive to the made of Im: J unp speed centrol as long as the steam generator level is kept within the 122nge t hat avoids protection or sdfeguardG QCluation.

Rod control 11 0 credit is taken for the operation of t he Rod Control rjstem for this tr.nsient, which results in an incr eare in L.S temperatute. With the Rod Control System in automatic, the control rods would cause a negative reactivity addition as thei ar e inserted in an at t.en pt to maintain RCS t e!rporatur e at its nominal value.

TJrblne Centt01 The turbine is modeled in the lead control mode, which is described in section 3.2.5.1 of Reference 2.

Auxillarv Fredwat ;]; f AFW flow would be credit ed when the re.f ety analysis value of the low-low steam generator level setroint is reached. However, the pat alvet er of interent f or this t ransier.t has reached its limiting value before the apptopriate Technicel Specifica. ion response time delay has elapsed.

Therefore, no AFW is actually delivered to the steam generators, iurbine Trir, The reactor trip leads to a subsequent turbine trip.

7.2.2 offsite Dose Calcul$ tion Input Analysis 7.2.2.1 11oda]ization

, Due to the asymmetry of this transient a two-loop model, with a single loop and a triple Icop, is utilized for this analysis. The boundary c>nditions f or the int set steam generators are syncetric, enabling the une of the two loop model. The pressurizer modeling includes the use of the local conditions heat transfer option for the vessel conductors.

7.2.2.2 Initial Conditions d

Core Power Level high initial core power and a ponitive uncertainty ma . :uize the primary system heat load.

freesuriner Presenig High initial pr es ure and a posit ive uncertainty delays the time of r eactor trip. This retards the primary system cooldown, extending primaty-to-secondary leakage, and therefore maximizing the offsite dose.

7-6

Prezrurizer L1"21 High initial level with a positive uncertainty maximizes primary-to-secondary leakage and tr.aximi zes pr essuri zer heater operat ter .

r

( Pe rtc t oi . 'A.;;r.<.tl._ Ave rau e Temne r at u r e Huminal vesrel a' ret age temper at utt_e with a negative t'ncertainty is used to minimize the initial st eam generat or st ecin pressule. This rttximizes the in2tial diftelential pressure across t . .e st eam generctor tubes and

! theref or e maximites the initial prin.ary-to-secondary leakago.

ECO FleM Nominel prinaty syntc:m loop flow with a negative uncertainty is assumed.

Low forced circulation flow .+esult) in lower natural circulation flow during the post-trip cooldown. This, reduces primary-to-secondary heat -

trancfer and extends plant cooldown.

Cor e Pn_ar s Flow Cot e bypass f J ow is not. an irnportant para neter f or this t ransient .

2 e m Ocnerator Level Minimum st eam generator level reduces the initial secondarv 2 nvent ory available to mix with and diluts the primary-to-secondary .eakage.

Fuel Tomoeiatur.e High initial f uel temperature naximizes the stored energy whicn must be temoved during the post-trap natural circulation cooldown.

rteam cenerater Tube Pluacina A bounding high tube plugging level minimizes the 11..t ial at eam generator steam pressure and therefore maximizes the pressure 3

diffet- tal between the pritrary and secondary systems.

Boundary Conditions f irm ie Failure The single failure identified for maximizing offsite dose is the failure of the PORV on ' w ruptured steam generator to close. In this analysis, this valve remains open until operator action is taken to isolate the PORV. Per Reference 4, page 5-7, "The most litniting failure would be the loss of air supply or power which prevents actuation of the (PCRVc) from the main control room. The valves could be operated (locally) by manual action to correct for this single failure." This failure is ,

incorporated into the analysis as it prolongs the trens.ient, maxirnizing the primary-to-secondary leakage.

Eressurizer enferv valves The pressuriner code safety valves are not challenged during tre course of this transient.

7-7

__ ____ _-___ --- a

1,"r iine rMoo nl zc The main stenn code safety valve 1 are modeled wit h lit t, accumulation, and blowdown a.aumptions which miriimize secondary piercure and maximize at tnosther ic ct eam r eleases df ite Pewn Offsite power in assumed t o be lost coinrident with turbine tiip This isolat es steam flow t o t he conderner , t her eby maximi zing t he atmosp heric steam teleares, n euk Madel

'Ihe br eak is assumed to be a double-ended guillotine break of a cingle stea.m generatot tube at the tubecheet curface on the steam generator out let plenum. This location maximizes the mass flow through the break.

12 , h raticn The reactor coolant pumps are assumed to operate normally until offsite powet is lost coincident with turbine trip.

M: infecticn SI actuation is assumed to occur on low pressurizer pressure at a r,etpoint with an applied positive uncertainty. Maxarnum ECCS injection flow is assumed to maximize t he primary-t o-secondary leakage.

M_ tin reedwat er Main f eedwater flow is assumed to terminate coincident with the loss of of f sit e power to minimize the secondary inventory available to mix witn and dilute primary-to-secondary leakage.

Char 11rn Flow A conservatively high charging flow capacity is modeled to delay teactor trip and maximize total primary-to-secondary leakage.

p 4 Manual A;t ions Immediate action to maximize charging flow (penalty).

Immediat a action to energine pressurine heater banks (penalty).

Identif y ano isolate ruptured steam generator consistent with assumptions in WCAP-10698 (Reference 5), 10 minute minimum delay (credit).

'solate the steam supply to the turbine-driven auxiliary feedwater pump from the ruptured steam generator after identification of the ruptured steam nenerator. An operator action delay time of 5 minutes is assumed (credit).

Isolate f ailed opm steam line PORV en the ruptured steam generator with an operator action deluy time trom when it should hc closed normally. The delay times assumed are 5 minutes for control room and 15 minutes for local operation (credit).

Manually control auxiliary feedwater to maintain zero power steam generator levels (ncminal),

a 7-8

L's j ng t he st eam line PCRVs , initiate natural circulation cooldown of the prinary syst e:n af ter identification of the rupautei ct uri genetatot. CI erat or act Ion doloy t iroes of L minutes Ior contIol roon act ion and 10 minuter fcr local action are assu:'ed (credit).

I n i t i at.e deptestutization of the 1 ritr ary syst em using t he r ossutizet i ORVs t o t tIminat e bteak ! low 2 tinuter affer the prinaty nystem in cubcooled at the ruptured st r am Jenerat or pressule, or 2 minutet after the cocidown has 1+en ct g.lete.1 teredit).

Terminate cafety injection flow when pressurizet level recovers with a conservative delay (penalty).

Manually control chaiging flow aftet safety injection termination to maint ain the reto power pressurizer level (nominal!

I 7.?.2.4 control, Protection, and Safeguards Syste.-m Modeling Yertor Trip A r eact or trip occurs on eit her low pressurizer pr essur e or overtemperature AT. The Technical Specification tesponse times are taed and the safety analys s setroints include the effects of uncertainty in the monitoreu parameter and in the setpoint Eterruri;er Precrure centr;.A This control syntcm is assumod to be in manual and therefore is not modeled. Operator action i< assumed to energize the precourizer heaters and control the PORVs. Pr essurizer cpray is not available for the duration of this transient.

Piersurizer level cent rol This control system is assumed to be in manual and ther ef ore is not modeled. Opetator action is assumed to maximize charging flow, rt <an Line POPvs and condenrer St en_Durp Th*# main stAtm PORVs are assumed to be operable for this transient with a positive bias applied to the control signal. This assumptien minimizes secondary pressure and maximizes atmospheric steam releases.

The condenset steam dump valves are not assumed to be operable.

Crndenser steam dump would nonconservatively minimize offsite doses.

S t e im Generat or Level centrol This control system is assumed to operate to maintain the initial steam generator level prior to reactor trip.

Mai n Feedwat ei Pu ri, roeed control This cont t ol system is assumed to operate to maintain the initial steam generator level prior to reactor trip.

7-9

__ ___ _ _ _ _____ _ _ _ _ ____ _ _ _ m

'hn bino cont r ol The turbine is modeled in the 1 cad control mode, which is described in Section 3.2.5.1 of Reference 2.

Auxilfar'* Feedwatri Auxiliary foedwater initiation occurs after the loss of offsite power with a delay, consistent with Technical Epec1*1 cations. A purge volun.e of hot water is assumed to be delivered before cold feeJwater reaches the steam generators. Minimum flow rates are assumed to rainimize priraary-to-secondary heat t ransf er .

.I MCIV Clo9ure Autonestic MSIV closure is assumed using both a dynamically compensated and a static steam line pressure signal. Early cicsure maximizes the pritriary leakage released to the atmosphere through the tailed open steata i lino PORV. I I

7-10 L.....

8.O r.tMMARY The pr eceding chaptet s have deset ibed in detail the systern analysis snodoling assunrt ions used by Duke tower Conpany f or the PSAR Chc.pter 15 it rident ana ly s tis not docu:nented in Keierence 1. Table 8-1 sun:1.ar i ne n these niodeling det ails for each of the analyzed events.

~

8-1

_ . _ _ _ _ _ _ - )

W Table 8-1 Accident Analysis Assumptions 15.2.7 15.2.'

FSAR Section 15.1.2 l 15.1.3 15.2.3 '

15.2.3 15.2.6 15.2.6 l.15.2.6 3.3.

7 3.'.?

Foport Section 2.2 l 2.3 3.1.1 3.1.2 3.2.1 3.2.2 } 3.2.3 3.3.1 I;cmina l Nominal High High High High Not:Inal High High t;ominal jower **

Pzr Pressure !2cminal Ncminal High High High **

Nominal High TJorinal P2r Level High High High High High **

Low High **

Lew RCS T+cp t;ominal IJominal High High High High Ic=inal High High Nminal FCa Flow TJaminal t!cminal Lcw High Lcw High Nominal Low High Nominal l

l Bypass Flow Hominal Nominal ** ** ** **

Nominal ** **

Nominal!

SG Lovol Low ** High High ** High ** High High High l Fuel Terrp Low Low LW Low Low Lcw High Low Low High I SG Tube Plugging None None High Nene High None **

High f.ono High P r Spray - -

Off Auto Cft **

Auto off Auto ,

Pzr Heaters Off Of1 Auto Auto Auto **

Off Auto i Oi!

Pzr PCRVs - -

Closed Auto Closed **

Auto Closod Auto SM F< V s Closed C1csad Closod Closed Cicsod Closed i Clo wd Clened m - -

8 Sto ,1 Closed Closed Closed Closed closed Clcsod I c csod Clorad to ** *

• _ - -

gg L,. u l MFX o ump Spood Pod Contrcl *

• Manual Manual - - -

Manual !M nua l Manual Auto Auto <* ** **

• Auto Auto Auto Turbino Control - -

SI Signal - - - - - -

SI Flow - - - - - -

S1 Dolay - - - - - -

LCSP LOSP LCSF A Lvl , LVI FG L'.* 1 l AFW Signal SG Lvl AF/ Flow Min ** *

• Min Min Min Min Min Min I AFX Delay TS ** *

• TS "' S TS TS T' T' Turb Trip Signal SG Lvl -- - -

LCSP LOSP LCSP F2 Trip Ex Trip Fx Trip Turb Trip Doiay TS -

Neno I;cno tiene None  !.ane Mcno U r nsa IF mo stm Line Icol - - - - -

Signal -

Stm Line Isol - - - - -

Dolay -

grT" Isol Signal SC Lvl -

l MFW I ol Delay TS - - - - - - '

I l

Notes:

Refer to the text discussion of this transient.

• • Fesults of the transient are insensitive to the choice about this parameter.

Not applicable, eithel because the transient does not challenge that contol system er because t he l malfunction of that system might be the cause of the transient.

Table 8-1 (continued)

Accident Analysis Assu:rptions FSAR Section 15.2.8 15.2.8 15.3.1 15.3.2 15.3.3 15.3.3

• 15.4.1 15.4.1 15.4.2 15.4.2 Saction 3.4.1 3.4.2 4.1 4.2 4.3.1 4.3.2 5.1.1 5.1.2 5.2.1 5.2.2 Power flominal High  !!oninal tiominal High High 0 fiominal [ I;cminal PZr Pressure Noninal Low flominal Nominal High Low , High Ziominal High lNcninal Low Low Low Low High Low High Lov High Low P r Level **

High **

IIomina l t;cr ina l RCS Temp IIcminal High Nominal Nominal High

• • **  !!cminal RCS Flow f;cminal Low Nominal Nominal Low Low Nominal Byrcass Flow 'Jomina l ** Nominal Nominal High High **

IIoni nal Nominal

} ** ** ** ** ** ** High High SG Level **

Lew Fuel Ten p High High High High Low High Low Low Low LOW SG Tube Plugging ** High ** ** ** ** High None High High

-Pzr Spray Auto Auto Auto Auto Off Alto 01f Auto Ot ,

• • ** Cff P2r Heaters off Off Off Off Auto OfI Off Pzr PORVs Auto closed Auto Auto Closed Auto Closed Auto Closed SM PGRV3 Closed Closed Closod Closed Closed Closed Closed Closed Clcsod Clo wl ca 8 Steam Dump Closed C1csod Closed Closed Closed closed closed C lowd Closed Cicsod

" ^

Auto Auto SG Level - -

MF"4 Pump Speed - -

Rod Control Manual Manual Manual Manual Manual Manual - - - -

Auto Auto Auto Auto Auto - -

Auto Auto Turbine Control Auto I - -

SI Signal - - - -

SI Flow -

Min - - - -

TS - -

SI Delay - - - -

AF'd Signal SI *

• AF"4 Flow Min -. . .

.. 7g gyg pglay Rx Trip Fx Trip Fx Trip Rx Trip Ex Trip Rx Trip - -

Py Trip RY Trip Turb Trip Signal None None None None !1ono - -

Nono Nero Turb Trip Delay None -

Stm Line Isol - - -

Signal - -

Stm Lina Isol - - -

Delay ** ** **

MFW Isol Signal -

Me"d Isol Delay -

l Notes:

Refer to the text discussion of t his transient .

• • Posults of the transient are insensitive to the choice about this parameter.

- f Jot applicable, either because the transient does not challenge that cont ol systum or because the n:al f unct ion of that s'/ stem might be the cause of the transient.

t_ . ______ - l

a . _-

y we < - l .

a . i Table 8-1 (continued)

Accident Analysis Assumptions FSAR Section 15.4.3c 15.4.3d 15.4.4 15.4.5 15.4.7 15.5 1 15.6.1 15.6.3 l 15.6.3 Section 5.3 5.4 S.5 5.6 5.7 6.1 7.1 7.2.1 i 7.2.2 Power -

Nominal

• l lNcminal Nominal Nominal High

' High P r Pressuro -

Nominal Low - -

Nominal f;cmina l Nominal P r Levol -

High Low - - High High Low High RCS Temp -

Nominal High - -

rJominal. Nominal Ncminal Lew PCS Flow -

Nominal Low - -

Nominal Nominal Nominal Lcw Bypass Flow -

Nominal High - -

Nominal Nominal Nominal SG Lovel -

High ** - -

Low Fuel Temo -

Low Low - -

High High High High

• • ** ** ** High SG Tube Plugging -

High - -

P r Spray -

Auto Auto - -

Auto -

Auto Ott P r Hoaters -

Off Off - -

Cff Ctf Oft Manual P :* r PCRVs -

Closed Auto - -

Auto -

Auto Manual SM PCRVG Closed -

Clcsod Closed Closal m - - -

8 Stoam Dump -

Closed - - -

Close<1 Closed Cicsod Clewd a Auto Auto

'elt o SG Level - - -

MFW Pump Spoed -

• • Auto - - Aut o Rod Control - -

Auto - -

Manual Auto Manual Auto Manual Auto 'uto Auto Auto Turbine Control - - -

i Low Pr l SI Signal - - - - -

Pross SI Flow - - - - -

Max -

thX CI Dolay - - - - -

Nono AFI Signal -

• • LO.c F

• ** ** Min AFW Flow - - -

Apw polay .

TS Turb Trip Signal --

) RX Trip - - -

Ex Trip RY Trip Ex Trip Fx Trip Turb Trip Delay -

l Mone - -

Nono None Non'> None Stm Line Isol - - - -

Signal -

Stm Line Isol - - - -

Delay ** **

MFW Isol Signal -

• • - - - - LC'4 MFW Isol Delay -

NOr:e _

Notes:

Refer to the text discussion of this transient. i

• • Fesults of the transient are insensitive to tha choice about this par roter,

- Not applicable, either because the transaer.t does not challenge that contol system or bocau w tho malfuncticn of that system might be the cause of the transient.

9.0 h EFERE!K'EL

1. Dl'C -lie - 3 0 01, *11 ult idimensional React or Tr ar.sierit s ard M!et y Analysis 111y t: 1cs Parametett Methodology", Duke s' owe r Crq 2ny, Fevision 1, Jurw, 1991.

2. Ol'C -!iE - 3 0 0 0, " The i ta l - ily d r au l i c Transient Analycir !!ot hodology " ,

Du r.e 1ower Company, Revision 2, Fe.bruary 20, 1990.

3.  !!ukcG-osco, av.t .!J.R.c. of f ice of 1 uclear Peactot hogulation standard Feview Plan", P uvi r. i cti 3, July 1981.

4.  !!UREG - 0 4 2 2 , SER Relat ed to t he Operation of McGuir e !Jucle at Ct at iota, Unitr4 1 ar.d 2, Euppleinent 4, January 1981. -

C. WCAI'- 10 6 9 8 - P - A .

• CGTR Analysis Methodology to Det errnine t he Italgin to Eteafn Generator Overf all", 8.11. Lewis, et.dl., August, 1987.

9-1

. e .,

't l' DUKE POWER 1

November 5,1991 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555

Subject:

McGuire Nuclear Station Docket Numbers 50-369 and -370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Oconee Nuclear Station Docket Numbcrs 50-269, -270, and -287 Final Respor se to Questions Regarding the Topical Reports Associated with the MIC8 Reload Package

References:

1) Lottar, H. B. Tucker to NRC, January 9, 1989.

(DPC-NE-2004 submittal)

2) Letter, H. B. Tucker to NRC, September 29, 1987.

(DPC-NE-3000 submittal)

3) Letter, H. B. Tucker to NRC, January 29, 1990.

(DPC-NE-3001 submittal) -

4) Letter, M. S. Tuckman to NRC, September 25, 1991.

(Reaf firmation of Proprietary Af fidavit for DPC-NE-2004)

5) Letter, M. S. Tuckman to NRC, September 25, 1991.

(Reaf firmation of Proprietary Af fidavit for DPC-NE-3000)

On October 7 and 8, 1991, representati q3 of Duke Powar met with NRC Staff and contract reviewers to diduuss outstanding issues associated with three Topical Reports (References 1, 2, and 3),

which are currently undergoing review. At this meeting, and during various telephone conference calls subsequent to the meeting, questions were identified which required additional information or clarification. Attached are formal responses to each of the questions. The attached infortation should resolve all outstanding issues related to the review of Topical Reports DPC-NE-2004, -3001, and -3000.

Please note that some of the information is identified as

__ _ - - - - - - - - - - - - - - - . - _ - - - - - - - _ - - - -- - - - - - - - J

4 Nuclear Regulatory Commission 1991 November 5, Pace 2 proprietary, and should be withheld f rom public disclosure pursuant to 10 CFR2.790. Affidavits attesting to the proprietary nature of the information have been provided (References 3, 4, and 5).

Also, please note that while aspects of the referenced Topical Reports may be applicable to all three of Duke's nuclear stations, approval of the Reports is required for McGuire Unit 1 Cycle 8; current 1j scheduled for startup in early December, 1991.

If there are any questions, please call Scott Gewehr at (704) 373-7581.

. Very truly yours, ,

s- -,

cw .,.,r r / y. . ,a:,, ,;.L

/

H. B. Tucker cc: Mr. T. A. Reed, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. 20555 Mr. R. E. Martin, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Washington, D. C. 20555 Mr. L. l. . Wiens, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 14H25, OWFN Mr. S. D. Ebneter, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323

U. S. Nuclear Hegulatory Commission November 5, 1991 Page 3 bxc (w/o attachments):

R. L. Gill, Jr.

P. F. Guill S. G. Denosole P. J. North G. B. Swindlehurst K. S. Canady R. II . Clark R. M. Gribble _

GS-801.01 i

,.-.A

l I

ATTACHMENTS Attachment 1: A discussion of the adequacy of the McGuire/ Catawba steam generator modeling in DPC-NE-3000 with respect to conservative prediction of primery-to-secondary heat transfer for transients which involve U-tubo l uncovery.

1 Attachment 2: Responses to informal questions on DPC-NE-3002, as j understood by Duke Power, regarding issues that were I not adequately addressed at the meeting, that were 1 requested by the NRC to be f ormally docketed, or thr.

arose in subsequent telephone convarr:ticut.

Attachment 3: Responses to informal questions on Chapter 15 markups, as understood by Duke Power, regarding issues that were not adequately addressed at the meeting, that were requested by the NRC to be formally docketed, or that arose in subsequent telephone conversations.

5.ttachment 4: A response to an additional question on DPC-NE-2004.

Attachment 5: A set of markups to DPC-NE-3001; due to questions asked at the meeting, and other corrections.

Attachment 6: A set of markups to DPC-NE '902: due to questions

. asked at the meeting, and it?er ccrrections.

+ ;- - - - _ _ - - -

,7 Attaciunent 2 Qxstions op %pical Report DPC-NF-3002 s g 1. Justify that the results of ans y/ir.g the feedwater flow increase transient at zero power will not be more IP iting than the full power case, considering the incrc- ' peaking factors, reduc 1 RCS flov ' asymmetric inlet temperature and power disir.Suu ts as a result of r. stuck rod

- r s' .oop malfunction.

"er increase due to a feedwater ' low inc.cese at zero power wou!c s worst.

=

-utron flux v actor trip low setpdat. This actpoint is no higher tha. 33M RTP.

.; r~ ctron attenuation in cooler downcomer water would af fect only the excore aux

'g W.teck , ,+

to em die alf cted loop (ai lea,t two detectors must indicate above the setpoint for a trip

. an allowance fo. this is already included in the 103 RTP margin between te --

1. v m analysis value and the 2M RT P Technical Specification trip setpoir t. For the core

4 powe: 4 ms at the stan of the accident (those permitted by the Technical Specincations),

aboulani b a macgin exists at 359 RTP wd. three (see proposed revision to Technical  %

Sperdi ut i~ ' 4.1.2) re:ctor coolant pumps operating. Unlike the steam line break accident.

W" '

a,L .juate a vn margin is maintained. Power generation :

• ccms when the contml raA fall inte ; core. The presence of a stuck nx' . .~. urb the core p7wer distabution dur'ng the time of minimum DNBR v . socal DNB margin due to a shift in core power distribution toward the quadi d loop would be somewhat mitigated by a gain in DNB margin due to reduc .6 inlet temperatures in that quadrant, it thould also be noted that the et ent justilication presented in the McGuire and Catawba FS ARs ,

for no 'yting J. tis case, Giat the reactivity insertion rate for this trans?.nt is less than the rate asvimed in the ut. cont:olled RoCA bank withdrewal f:Om zero power, remains valid for the ai Duke Power Compay w % , rproach.

4

2. For the excessive iner,se in sceondary 3 team llow accident, explain why a low initial pressuriter bubble volume would rwximize the pressure decrease due to contraction. g Respe: When the coolant contracts the pressurizer steata bubb!c expands, reducing the 4

pres 3eiter pressure. 'lhe amount of pressure reduction is roughly inversely proponional to the volume of the bubble. Therefore expansion of a smaller bubble will mnimize the resulting pressure decrease.

M 3. For tne turbine trip accident, discuss the n:asons why a DNB analysis is not ret}uired.

Response: The FS AR will be revised to insen the following paragraph into the 15.2.3.2 " Method of Analysis" settlan:

For the turbine trip event $c reactor power, the core power distribution, and the core flow change very little prior to reactor trip. The RCS pressurization due to the reduction in secondary heat sink more than offsets the increase in core inlet temperature. Therefore significi.at DNB margin is r'aintained throughout the transient , and no quantitative DNB analysis is req iired.

?

l Nb  :

4. Explain why a high initial sicam generator level would maximite the transient secondary pressure response for the turhine tnp peak RCS pressure and peak secondary pressure analyses.

Response; initial steam generator (SG) level has a small impact on two competing phenomena during a turbine inp event. First, as secondary pressure increaser,ine saturated liquid in the SG becomes slightly subcooled, w hich causes some of the energy transferred from the primary to

)

heat the SG liquid to saturation. He higher the initial SG level, the greater the mass that is s subcealed during pressurization, which ends to cause a lower transient SG pressure. %c second pienomenon is that the higher the initial SG level, the smaller the initial steam volume, which tends to cause a higher transient SG pressure.

For the peak secondary pressure analysis, a sensitivity study on initial SG Iml in the turbine trip event was perfonned in a previous analysis. De sensitivity study demonstraad that a decrease of 8M yan from nominal for initial SG level resulted in a decrease of 0.15 psi in peak secondary pressure his result demonstrates that the effect of reducing initial steam volume dominates the cflect of increasing initial SG mass, and high initial SG level is conservative. In addition, this result demonstrates that initial SG level is not a critical parameter. The effcci of initial SG level is dominated by other conservatisms such as the drift and accumulation assumptions of the main steam safety valves.

For the peak RCS pressure analysis, the impact of initial SG level on the peak primary pressure j analysis is hmited to the second order effect of SG level on primary-to-secondary heat transfer.

Maximum primary pressur.: is achieved by minimizing pnmary-to-secondary heat transfer.

Primary-to-secondary her.1 transfer is less with an increase in the secondary saturation f

temperature, which increases with an increase in secondary pressure. Therefore, initial SG level i is chosen to maximi /c secondary pressure. As shown above, maximum secondary pressure ; l achieved with a high initial SG level. In addition, the effect of initiat SG level on secondary  !

pressure is dominated by other conservatisms such as the drift and accumulation assumptions of the main steam safety valves and the pressuriter safety valves.

S. Foi a given high pressuriter pressure reactor trip setpoint, justify that a lower initial indicated pressuriter pressure,i.e., one at the limit of Technical Specification 3.2.5.b, would not give a higher peak RCS pressure result for the turbine trip and uncontrolled RCCA bank withdrawa; ai power events.

Response: Wh!* it is tme that Technical Specification 3.5.2.b permits pressmizer presse e to indicate as low as 8 5 psi tal McGuire)less than the nominal value, this does not occur during automatic pressure control. Dere is no normal operator evoludon during manual control at luwer operation w hich reduces pressure belxx the nominal value. Therefo' this initial condition is not regarded as a credible one. Nevertheless,9ere remains suhicient margin in the pressure initial condition uncertainty adjustment to compensatt for this 8.5 psi. Therefore the results presented in the McGuire 1 Cycle 8 reload submittal FSAR markups are conservative. In addition, for the uncontroLd RCCA bank withdrawal at power, the initial margin to trip is 45 psi more than for the turbine trip event, thus ensuring further conservatism. The values used in the analyses are as follows:

I

)

~

r Parameter Turbine Trip UCBW @ Power Pressure lustrument Uncertainty, psi 20 20 Pressure Uncertainty Allowance, psi 30 45 Initial Actual Pressure, psia 2280 2250 Initial Indicated Pressure, psia 2250 2250 Trip Setpoint, psia 2400 2445 initial Margin to Trip, psi 150 195 Actual Pressure at Trip, psia 2430 2445

6. One acceptance criterion for the Condition 11 events .s that there should l , water relene from the pressuriter safety valves. For the loss of offsite pi wer, loss of nonu reedwater, and uncontrolled RCCA bank withdrawal at power accidents, provide analysis assumptions and results for widrecsing this criterion or justify that the margin to a water solid pressurizer r condition for Wese events could be bounded by another event (s).

Response: Pressuriier overlill is a poteatial concem during an event in which either safety injection (SI) occurs o' RCS heatup due to primary / secondary power mismatch occurs. As shown below Si is the key factor in pressurizer overfill.

Pressuriier level at the stuion is determined by the difference in pressure between two elevations a the pressurizer. In addition to the random cliccts typically associated 'vith a measuremeat, the use of a DP transmitter to determine pressurizer level introduces the possibility that a difference between actual liquid density and the calibrated density could allow actual pressurizer level to be higher than indicated level. This situation occurs when die actual liquid density is less than the liquid density at calibration conditions. Liquid density is a weak function ot liquid pressure, but liquid density is a strong function ofliquid temperature. TP mfore, actual livid density less

.han calibrated liquid density only occvs when actual liquid temperature is e eater than the liquid temperature at calitwaan. The pressurizer leve; transmitters at the station are caliNated at full power conditions. Since the pressuri/cr is at saturated conditions at calibration, tk _

lemperature of the liquid in the pressurica at calibration is the saturation temperature at nominal _

pressure,2250 psia, which is app oumately 653 SF, in order for the density error to cause actual level to tie greater than indicated level, the water entering the pressuriter during a transient must ,

be greater than 653 F, but no transient will achieve this hot leg temperature for a sufficient duration prior to mitigating actions occurring. In addition, since the initial hot leg temperature is less than 653 F, the initial insurge will decrease the temperature of the pressuriier liquid, /

causing indicated level to be higher than actual level. Pressuriter level in the McGuire/ Catawba 3 RETRAN model is not currendy determined by the difference in pressure between two elevations in the pressurizer. As stated in Section 3.2.4.1 of DPC-NE,3000, pressurizer level in this model is determined directly from the liquid volume actually calculated by RETRAN in the node representing the pressuriier. This modeling determines the actual level in the pressuriter during the simulation, and it is not possible for the calculated indicated level to differ from the calculated actual !cvel by more than the uneenainty allowance, and no credit is taken for the density effect described above Therefore, pressuriier level derived from cidier DP or liquid lewl will prevent a pressuriier overfill condition prior to reactor trip.

In tne loss of offsite power event (LOOP), Si does not occ r. Reactor trip occurs at the initiation of the transient, and no significant post trip degradation of the secondary side cooling ability reiative to the pnmary power generation occurs which could cause a power mismatch and

)

subsequent pressuri/cr overfill Derefore, since Si does not occur, and a signincant power misma th does not occur, pressuruer overfill does not occur for the loss of offsite power event.

]

In the loss of nonnal feedwater (LOFW) event, Si does not occur. Pressuri/er overfill could not o

' ccur prios to reactor trip because of the high pressuriier level reactor trip function. 0" dill is most severe in the Condition IV feedwater line break (FWLB) event, and the high prtssuri/cr level in the this event is caused by the addition of SI water to the RCS. Analysis has shown that continued Si causes pressurifer overfill to occur in the FWLB cvent. In compar: son, the LOFW transient is the most severe intact steam generator tube uncovery event, but the power mismatch is not suf ficient to cause pressuriter overfill. Werefore, since Si does not occur, and the power mismatch is not sulliciently severe, presurizer overfill does not occur for the loss of normal l feedwater event, in the uncontrolled RCCA bank withdrawal at power event, SI does not occur. Pressurizer overfill could not occur prior to reactor trip becau of the high pressuriter level reactor trh function. Here is no post trip degradation of the secondary side cooling ability relative to the primary power generation which could cause a power mismatch and subsequent pressuri/cr i overfill. %crefore, since Si does not occur, and a significant power mismatch does not occur, p rssurizer overfill does not occur for the uncontrolled bark withdrawal at power event.

7. For the loss of offsite power, loss of nomial feedwater, and feedwater line break accidents, provide a description of the auxiliary flow assumptions used, including number of pumps assumed. capacity, and flow fraction delivered to each steam generator.

Response: We limiting single failure in the Auxiliary Feedwater (AFW) System is assumed for each of these accidents. His assumption will result in no credit being taken for the single AFW pump whose loss would represent the greatest reduction in flow delivered to the intact steam generators nerefore only two of the three AFW pumps are assumed to be operating for any of these acciden' . Pump capacity is conservatively reduced from the manufacturer's head curves.

He reduced capacity corresponds to a pump performance level below that which the pump is verified to meet in periodic !csts. His additional reduction provides margin for further pump degradation between tests. The flow fraction delivered to each steam generator is calculated based on a model of the AFW pumps and piping layout. He flow fractions vary with 1) the transient backpressure in the steam generators 2) which station (McGuire or Catawba)is being analyzed. and 3) whether operator action has occurred to realign the AFW System to change which pumps deliver flow to which steam generators. Because of these variabilities, a separate time dependent AFW boundary condition is calculated for each plant for each accident.

8. Justify why only the double-ended feedwater line b cak is analyzed and not a spectmm of feedwater line breaks. A smaller break, for which the reactor trip occurred on low-low steam '

- generator water level might be limiting compared to the double-ended rupture of the main feedwater line.

Response: The current McGuire FSAR states (p.15.2-15)," ..it has been shown that the most limiting feedwater line ruptures are the double-ended rupture of the largest feedwater line..."

His assumption was reviewed and apwoved in the NRC SER for initial startup of McGuire.

Duke did not analyze a spectrum of break sizes since it was apparent that the issue of break size

~

was resolved per the existing FSAR analyses. Ilowever, in order to res[xmd to the question, additional investigation into the technical basis f or concluding 6at the double-ended rupture is the limiting break si/c was performed. The results of this investigation have led to the conclusion that the double-ended mpture is the limiting case.1he bases for this conclusion are as foilows.

The feedwater line break transient causes a reactor trip on either low-low steam generator level or high contaimnent pressure (~1 psig). The pre-trip transient response is dictated by the break sue.

For large breaks, the afIceted steem pencrator rapidly blows down, and main feedwater now to all four steam generators is lost out the break. 'the intact steam generators g.adually boil off until auxiliary feedwater delivery begins approximately 60 seconds after reactor trip. l.ong-tenu decay heat removal is established via auxiliary feedwater and the intact steam generators. For smaller feedwater line breaks, the alicued steam generator blows down more slowly and some main feefwater flow continues to be delivered to the intact steam generators. Main feedwater -

would only be stopped by assuming a loss of of fsite power coincident with reactor trip. For all break sizes, the affceted steam pencrator will blow down to dryout, and will not contribute to long-tenu decay heat removal. The minimum inventory in the intact steam generators is the key parameter when detemiining the linuting break si/c. The inventory in the intact steam generators will change depending on the pre-trip steaming duration, and main feedwater flowrate. Both of g these are a function of the break size. For larger breaks the steaming duration is short due to a rapid reactor trip. Larger breaks will also prevcot any main feedwater flow from reaching the intact steam generators. For small breaks the sicaming duration is longer, but some main ,

feedwater can still reach the intact steam generators. From this argument it follo >s that an intennediate si/c bn'ak will result in the minimum intact steam generator inventory. D is noted that the auxiliary feedw ater flowrate is the same for all break sizes. Therefore, the long tenn cooling capability is not affected by the break size. The break site concem is limited to  ;

detennining if the minin'um post-hip heat sink, carresponding to the minimum intact SG inventory case, causes any of the acceptance criteria to be met.

A sensitivity stud model were { y on break site was perfonned. The requirca modifications [

llhe effect of these modeling changes is to conservatively predict the minimum main feedw ater flow delivered to the intact steam generators. As stated above, for the double-ended rupture this flow will be

/cm, but as the break site decreases some flow will be delivered. All main feedwater is assumed to be lost on reactor trip due to an assumed kss of offsite power, in addition to the deable-ended rupture break sue for which all main feedwater is lost, split breaks of 0.2387,0.3, and 0.5 ft' were analyzed. Due to the feedwater nottle th w restrictor area

( )the Juration of blowdown and the time of reactor trip is unaffected until 'he break size approaches the range of sites analy/cd. Within tb "ange of break si/cs, the larger sizes predict an calier reactor trip,less steaming from the intact steam generators, and less main feedwater reaching the generator. Smaller break sites predict a later reactor trip, more steaming from the intad steam generators, and more delivered 1.iain feedwater. The integrated effect of these parameters on the analysis is characteri/cd by the minimum intact steam generator inventory.

1

. j Ilreak Si/c Rx Trip Time Minimum SG (IF). (sec) Mass (thm/SG)

DE 19.5

- i 0.5 0.3 -

De 0.5 ft break 2

size is significant in that this brea. size is at the transition where all main feedwater is lost out the break. The 0.3 ft 2break has some main feedwater reaching the intact steam generators, which explains the higher minimum mass flow. For breaks larger than 0.5 ft2 .

there is essentially no difference when compared to the double-ended break. De difference in reactor trip time is[ ) second, and the minimum steam generator masses are {

}fireak si/cs above 0.5 ft are thc[,

3 Ilased on the results of the break size sensitivity, the FS AR statement that the double-ended rupture is the limiting case is confinned. The minimum steam generator inventory in the intact steam generaton; and the available auxiliary feedwater capacity ensum that the consequences of the limiting feedwater line break are acceptable.

1

9. The current FSAR analysis of the feedwaterline break assumes that the AFW flow to the l faulted steam generator is spilled $ rough the break. Justify why this AFW spillage is not assumed in the McGuire i Cycle 8 reload submittal analysis of feedwaterline break. Justify the operator action time assumed isolate the AFW llow to the faulted steam generator.

Resp (mse: The McGuire and Catawba steam generators are of the preheat design with a lower feedwater noule entering the preheater region and an upper nonJe entering the upper region of the downcomer. At all but the lowest nower levels, the vast majority of the main feedwater flow enters through the lower nozzle. At all conditions the AFW ilow entes through the upper nouJc. AFW spillage through a break of a pipe connected to the lower nonle (main feedwater line) would require failure of a check valve plus closure failure on a second valve which receives a feedwater isolation signal, This is not regarded as credible. Otherwise, the AFW flow entering the faulted steam generator must travel down the downcemer and partially up the tube bundle through the preheater to exit through the lower nozzle bef ore reaching the broken piping. This flow path allows heat to be transferred to this water from the RCS. Since this AFW flow to the depressurized steam generator is a relatively large fraction of the total AFW flow, isolation of it reduces the total AFW flow available for RCS heat removal. Therefore its isolation is conservatively accomplished at two minutes into the transient. This is a very short time for the  !

operator to perfonn the steps in the emergency procedures preceding isolation of AFW to a faulted steam genvator.

10. For a feedwater line break analysis in which reactor trip was actuated by the low-low steam generator water level, a higher initial steam genemtor water for the faulted steam generator could be more conservstive.

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! Response: Duke agreesi If the low-low steam generator narrow range level reactor trip is credited in the FSAR Section 15.2.8 feedwater line break accident, a high initial condition uncenainty adjustment will he applied to the faulted steam generator narrow range level;

11. Justify the gap heat transfer coefficient used for the panial and complete loss of forced flow events, f

Response: A low fuel gap conductivity, which corresponds to a high initial fuel temperature is used in die analyses. The low fuel gap conductivity has three major effects in these events.

fNst, low gap conductivity is conservative due to the higher initial stored energy in the fuel compared to the initial stored energy in the fuct with a high gap conductivity. For a given rate of '

decrease of the fuel rod surface heat flux, a nigher initial stored energy in the fuct causes a higher i

heat flux during the Inuisient because more energy is available to be traosferred to the coolant.

Second, low gap conductivity is more conservative because the energy wul be retained in the fuel for a longer period of time. Since the energy is retained in the fuel for a longer time period. the heat flux will decrease more slowly, and compared to high gap conductivity, the heat tiux will be maintained at a higher value during the flow coastJown. Third, from a reactivity insettion aspect, low gap conductivity is less conservative than high gap conductivity. The fuel Doppler temperatum coefficien mTC)is negative, and a low gap conductivity causes more fuel heatup to occur, wl...n ad6 more negative reactivity than a high gap conductivity. However, the fuel temperature increase is minimal in these events, and the difference in reactivity insertion due to a low gap. conductivity versus a high gap conductivity is insignificant. The higher initial stored

[ ' energy and the slower heat flux decrem completely dominate the reactivity effect, snd low gap l- conductivity is conservative.

12. The stall's peak RCS pressure acceptance criterion for the locked rotor event is 110% of the

• design pressure instead of the proposed 120% value.

Response: Die Standard Review Plan (NUREG-0800, July 1981) states diat the Reactor Systems Branch acccmucc criteria for both the feedwater line break and the kicked rotor events are based oo'.necting the relevant requirements of General Design Criteria 31 as it relates tc the reactor coolant system being designed with sufficient margin to ensure that the boundary behaves in a nonbrittle manner and that the probability of propagating fracture is minimized. Although the -

woniing is the same fbr tuth events, the feedwater line break Standard Review Plan section also--

gives quantitative acceptance criteria, i109 of design pressure for low probability events and

,120% of design pressure for very low probability lcvents. He h>cked rotor event is characterized as a Condition IV event (limiting fault not expected to occur during the life of the plant) in both the McGuire and Catawba FSARs. -De acceptance criterion of 120% of the design pressure is i based on' assuming that a locked rutor event is of very low probability and therefore that the -

L acceptance criterion adopted for doubic-ended guillotine feedwater line breaks applies. However,

the peak pressure result of the k>cked rotor event analy/ed in the McGuire 1 Cycle 8 reload nubmittal is within i10% of the Reactor Coolant System design pressure.

' 13. Discuss whether the locked rotor event will be analy/cd assuming coastdown of undamaged pmnps coincident with turbine trip if n is more limiting.

' Resportse: The relevant Standard Review Plan instruction conceming this question is as follows:

cwt---r'+gp w. +- g -

s-v r v" ' -- W q

'"Ihis event should be analyzed assuming turbine trip and coincident loss of of fsite power and coastdown of undamaged pumps."

As stated in Section 4.3.l.3 of Duke Power Company topical repon DPC-NE-3002, offsite power was assumed to be lost coincident with turbine trip. Upon the loss of offsite power, voltage and

' frequency legin to decay on the four UM) V busses w hich supply power to the reactor coolant pump (RCP) motors. Pump motor speed, and therefore pump flow, decrease as the bus f requency decays. Af ter the loss of olisite power the bus voltage decreases to the RCP  :

undervoltage trip setp> int. At this point the RCP motor breakers open and the pumps coae ilo,vn as controlled by the inertia of their flywheels. It is Duke Power Company's position that dtis modeling meets the intent of the Standard Review Plan, i.e., the RCP coastdown is caused by the loss of ofIsite p>wer.

14. Provide and justily the gap heat transfer cocincient used for the k>cked n> tor accident.

1 Response: See response to question i1 above.

55, Justify the use of a point kinetics model for the uncontrolled RCCA bank wididrawal from

/eni power %nd single RCCA withdrawal events. Describe in detail the medmdology used to detennine the lounding radial and axial power shapes and the uncertaintics assumed. Provide and justify the Nnnding radial and axial power shapes, moderatar density coefficients, and trip

~

reactivity used in the DNB analysis.

Resp >nse: The uncontrolled bank withdrawal from zero power and the single rod withdrawal events use a point kinetics model to detemiine the core average power response. The point model employs physics parameters that conservatively bound the con designs. Due to the absence of leakage and spatial effects in a point model relative to a 3 dimensional space-time model, the point model win overprediu reactivity and the transient core average power response.

Spatial effects are accounted for by explicit 3-dimensional simulation of the core nower distribution. The analytical methmlology is as follows. Ec system thennal-hydraulic analysis is 1 conservatively simulated with the RETRAN code. Transient core thermal-hydraulic boundary conditions from RETRAN arc dien input to VIPRE to detennine DNBR vs. time. The minimum I

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,,-.,v . , __, _ . . , _ . . - .-.. w _ ..

DNBR limit. Thc3e nxh are assumed to experience cladding f ailure and contribute to the source tenn for the of fsite dose calculations.

The moderator density coef ficient assumed in the rmalyses is consistent with the cu Tent Technical Specification value.

The conservatise nonnalized trip reactivity vs. nonnalized drop time given in Figure 15.0.5-3b of the McGuire 1 Cycle S reload submittal was used. A rod drop time consistent with Technical Specification 3.1.3A was assumed. As stated in Section 2.2 of Duke Power Company topical repon DPC-NE-3001, the minimi t was assumed for the amount of trip reactivity insened af ter trip. As described in that rei worth assumes that the most reactive rmi remains in the fully withdrawn pasition and uw the ither rods drop from their power dependent insertion limits. {

3

16. The current FSAR high neutron Dux trip setpoint uncertainties inchale a process measurement accuracy tenn for shielding effects and detector placemert. Discuss how this effect is accounted for in the uncontrolled RCCA bank withdrawal accidents.

17. Specify which of the accidents discussed in DPC-NE-3002 will be analy/cd with a non-icro value of the pressurizer interregion heat transfer coefficient.

Response: The methodology for determining whether to model interregion heat transfer was i presented in the NRC/ITS/ Duke Power meeting on October 7-8,1991. Of the transients discussed in DPC-NE-3Dn?, {

3

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- 18. Explain what differences, if any, are intended by the use of various descriptions ofinitial condition assumptions in the text and tables of DPC-NE-3002, both among the text sections for various accidents and between the text rid Table 8-1 for a panicular accident.

Response:'The entries in Table 8-1 of DPC-NE-3Od2, which are in a standantized format for specification of initial conditions, are intended to be consistent with the text descriptions. Any variation in text descriptions has no significance. In general , the choice of initial conditions is as

. described in the first full paragraph on page 12 of DPC-NE-3002. 'nie specific table entries and their intended meanings are as follows:

"Nonunal" means that, for power level, the initial condition is chosen from the range of rem to 100% RTP based on conservative modeling or on consistency with the Standar:1 Review Plan. For pressurizer pressure the initial condition is the plant reference pressure

. for power operation For reactor vessel average temperature the initial condition is the pmgrammed value for the chosen power level. For core bypass flow the initial condition

-is the best estimate calculated value. For RCS flow the initial condition is a chosen valac at or below the Technical Specification minimum measured flow, initial condition uncertainties in each of these parameters is accounted for in the statistical DNB limit, only these five parameters are initialized at nominai values, and the specifica ion of

" nominal"is used only for DNB analyses.

filigh" raeans that, for power level, the initial condition is chosen as for " nominal" and then increased by the ic;tial condition uncertainty. For pressurizer pressure, the above reference pressure is increased by tha initial condition uncertainty. For reactor vessel average temperature, pressu izer level, and steam generator level, the programmed value for the chosen (unadjusted) power level is increased by the respective initial condition

. unsertainties. For core bypass flow the initial condition is the best estimate calculated value increased by the assumed SCD uncertainty in bypass flow. For steam generator tube plugging the initial condition is a plugging level above that existing at any of the four,McGuire and Catawba units _. Foi fuel temperature, a conservatively high core average calculated value is used.

Low" means that,- for power level, pressurizer pressure, reactor vessel average temperature, pressurizer level, steam generator level, and core bypass llow, the initial condition is the same as for "High" except the initial condition uncertainty adjustment is a decrease rather than an increase. For fuel average temperature the initial emulition is a

nominal core average calculated value.

"None" is used only for sicam generator tube plugging and means zero plugged tubes.

, ' "**" mews that initial condition uncertainty is unimportant since the results of the transient are insw m to the exact value af the parameter.

" " means that the unlel used to analyze the transient did not have an explicit input for the panuneter in question L

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i (19. liigh steam generator level is etated to be conservative for peak RCS pressure for the turbine -

. trip peak primary pressure analysis. Low steam generator level is stated to be conservative for

peak RCS pressure for tbc loss of AC power transient. Explain this discrepancy.

Respmse: The statement in the loss of AC power transient discussion on p.3-7 is in error. It should be replaced with, " Initial steam generator level is not an imponant parameter in this analysis." Table 8-1 will also be corrected.

20f For the inadvenent operation of ECCS during p>wer operation transient, high steam generator tube plugging is assumed. Explain this assumption considering that steam generator level is stated as unimponant, and therefore that heat transfer must be unimponant. This

- discrep:mcy appears to occur in other transients.

Response: Steam generator tube plugging is unimponant for this transient.- The text on p.6-2 and Table M 1 will be revised accordingly. ' Based on this question additional review was cu. ducted of the text of DPC NE-3002 vs. Table 8-1. Marked-up pages are included in >

LAttachment 6. In general, these markups currect inconsistcacies between the text and table or in the use of table entries defined in the response to Question 18.

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