NOC-AE-000565, Forwards Amended Pages for Insertion Into South Texas Project Nuclear Operating Co Previously Submitted Response to NRC Rai.New Pages Include Expanded Answer to Question 4.b

Forwards Amended Pages for Insertion Into South Texas Project Nuclear Operating Co Previously Submitted Response to NRC Rai.New Pages Include Expanded Answer to Question 4.b

ML20212J011
Person / Time
Site:	South Texas
Issue date:	06/16/1999
From:	Thomas S HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NOC-AE-000565, NOC-AE-565, NUDOCS 9906280290
Download: ML20212J011 (14)

Text

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Jum 16,1999 NOC-AE-000565 File No.: G20.02.01 G21.02.01 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document ControlDesk Washington,DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 Amended Pages for Response to Request for AdditionalInformation on Proposed Amendment to Technical Specifications to Reflect Replacement Steam Generator Reactor Coolant Flow Differences

References:

1) Letter from S. E. Thomas to U.S. Nuclear Regulatory Commission dated May 20,1999, (NOC-AE-000540) " Response to Request for Additional Information on Proposed Amendment to Technical Specifications to Reflect Replacement Steam Generator Reactor Coolant Flow Differences" Attached are amended pages for insertion into South Texas Project Nuclear Operating Company's previously submitted response (Ref 1) to the U. S. Nuclear Regulatory Commission Request for AdditionalInformation. The new pages include an expanded answer to question 4.b.

(

Please replace pages 1 of 11 through 11 of 11 in NOC-AE-000540 with the attached pages 1 of 12 through 12 of 12. All text pages are replaced because of changes in page numbers and page mugins to accommodate the added detail. No other significant changes were made.

If there are questions, please contact Mr. M. E. Kanavos (512) 972-7181, or me (512) 972-7162.

/

') 9 i 4 S. E. Thomas Manager 'JD/

Design Engineering sw SET /MEK/MTVN oMk98 PDR STI: 30897184 J

~. ,

NOC-AE-000540 L

May 20,1999

l. .cc:

. Ellis W. Merschoff Jon C. Wood Regional Admmistrator, Region IV . Matthews & Branscomb U. S. Nuclear Regulatory Commission One Alamo Center 611 Ryan Plaza Drive, Suite 400 106 S. St. Mary's Street, Suite 700

, Arlington, TX 76011-8064 San Antonio, TX 78205-3692 1'

, Thomas W. Alexion Institute of Nuclear Power

' Project Manager, MailCode 13H3 Operations - Records Center U. S. Nuclear Regulatory Commission 700 Galleria Parkway Washington, DC 20555-0001 Atlanta, GA 30339-5957 Cornelius F. O'Keefe Richard A. Ratliff Sr. Resident Inspector Bureau of Radiation Control c/o U. S. Nuclear Regulatory Commission Texas Department of Health i P. O. Box 910 1100 West 49th Street Bay City, TX 77404-0910 Austin,TX 78756-3189 J. R., Newman, Esquire D. G. Tees /R. L. Balcom Morgan,I2wis & Bockius Houston Lighting & Power Co.

1800 M. Street, N.W. - P. O. Box 1700 L Washington, DC 20036-5869 Houston,TX 77251 M. T. Hardt/W. C. Gunst Central Power and Light Company City Public Service ATTN: G. E. Vaughn/C. A. Johnson

P.'O. Box 1771 P. O. Box 289, Mail Code: N5012 San Antonio,TX 78296 Wadsworth,TX 77483

- A. Ramirez/C. M. Canady U. S. Nuclear Regulatory Commission

City of Austin Attention: Document Control Desk-:

Electric Utility Department Washington, D.C. 20555-0001 721 Barton Springs Road Austin,TX 78704 l

l 5

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NOC-AE-000540 May 20,1999 l Page1of12 Request for Additional Information Regarding Technical Soecifications to Reflect  !

Replacement Steam Generator Reactor Coolant Flow Differences South Texas Pro _iect. Units 1 and 2 1.

Page 2 refers to two changes made to reduce vertical uplVtforces on the reactor core: (1) removal of the reactor vessel headflow nouleplugs (T-cold conversion), and (2) removing thimbleplugging devices on thefuelassemblies. Pleaseprovide or reference drawings showing each of these modifications.

(See annotated drawings in Attachment 1)

2. Discuss orprovide references to the methods that were used in determining the OTAT '

and OPAT trip setpoints. Justify that the safety limits and[OTAT and OPAT trip] \

setpoints are adequate to providefor DNBR protection. I As discussed in WCAP-8745-P-A (Ref.1), reactor core safety limits are comprised of vessel exit boiling limits and DNB core thermal limits (DNB limit lines). The core safety limits presented in the South Texas Project (STP) Technical Specifications (TS) are defined in terms of reactor coolant system Tivo as a function of power level. Because increased vessel

{

flow rate causes a downward shift in T Avo for a given Vessel inlet temperature, TS Safety '

Limits had to be revised. WCAP-8745-P-A (Ref.1) methodology was applied to STP Replacement Steam Generators and results showed that existing OPAT and OTAT setpoint equations protect core safety limits. Therefore, OTAT and OPAT setpoints provide adequate protection of core thermallimits.

3.

Page 6 states that a specified number of the reactor vessel upper headflow noule plugs will be removed to increase coolingflow to the upper head region so that the vesselinlet temperature (T-cold) will be achieved in this region. Discuss andjustify how this number willbe determined.

A thermal-hydraulic reactor internals vessel evaluation was performed to predict upper-head region fluid temperature by modeling the STP pressurized water reactor vessel and internals system. This approach to analysis has been justified by comparing calculated results with  !

empirical data from various plants. Excellent agreement between calculated results and actual measurements confirms that the methodology yields reliable information.

An iterative solution technique was used to establish the reactor internals modification design for directing flow from the downcomer region into the upper head region. This design process also considered uncertainties on the pressure loss coefficients in order to ensure that the vessel inlet temperature (T-cold) will be achieved in the upper head region at 100%

steady state full power STI: 30897184 s

. NOC-AE-000540 May 20,1999 l Page 2 of 12 4a. 4 What specific events were reanalyzed, what specific events were not reanalyzed, and providejustificationfor not reanalyzing the other events. l 15.1.1 Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature l

Event evaluated.

1 South Texas UFSAR (Section 15.1.1) discusses increased thermal load due to opening of the high pressure feedwater heater bypass valve, or closing of steam extraction valves to the high pressure feedwater heater. These would result in a transient very similar to the excessive load increase event (UFSAR 15.1.3), but one of reduced magnitude. This remains a valid conclusion with the A94 replacement steam generators (RSG). This feedwater temperature reduction event is also bounded by the feedwater malfunction event reanalyzed for the RSG program that causes an increase in feedwater flow (UFSAR 15.1.2).

15.1.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow e Full Power Conditions -

Event reanalyzed e Hot Zero Power Conditions - Event Evaluated With Model E cteam generators, results of the hot zero power (HZP) feedwater j

malfunction (FWM) analysis are non-limiting, i.e., no voids were predicted, thus there is no DNB. Compared to results of the HZP steamline break (SLB) analysis with Model E steam generators, results of the HZP FWM analysis are much less limiting. Based on changes expected with A94 replacement steam generators, e.g., reactor coolant flow differences, the HZP MSLB analysis j results bound results of HZP FWM analysis.

{

15.1.5 Spectrum of Steam System Piping Failures Inside and Outside Containment l

Hot Zero Power Conditions - Event reanalyzed Results of HZP MSLB - Core Response analysis for the replacement using {

Model A94 steam generators (RETRAN) are much less limiting than results i obtained in analysis with the Model E steam generators (LOFTRAN).

• Full Power Conditions -

Event reanalyzed j

Although results of the full power MSLB analysis are not discussed in the South Texas Project UFSAR, a cycle-specific DNB evaluation is performed for every STPNOC core reload design.

STI: 30897184 w

'* NOC-AE-000540 May 20,1999 Page 3 of 12 15.2.2 Loss of External Electrical Load -

Event evuluated As discussed in the South Texas UFSAR, a loss of external load event results in a Nuclear Steam Supply System (NSSS) transient that is less severe than a turbine trip event. Thus, a detailed transient analysis is not performed for this event because it is bounded by the reanalyzed turbine trip event.

15.23 Turbine Trip -

Event reanalyzed 4b.

Provide values for the partial derivatives used in the RETRAN DNBR model andjustify that these values are conservativefor the DNBR analysis ofSouth Texas.

Figure 1 shows two sets of core thermal limit lines in terms of core inlet temperature as a function of core power for the South Texas Project (STP). The limit lines identified as

" Actual" were developed based on a reference power shape, using the methodology described in WCAP-11397-P-A (" Revised Thermal Design Procedure") and were used to generate Technical Specification Figure 2.1-2 and to revise Updated Final Safety Analysis Report (UFSAR) Figure 15.0-1. These " Actual" limit lines are part of the reactor core safety limits which, as discussed in the response to question 2 of this request for additional information (RAI), were shown to be protected by the existing OPAT and OTAT setpoint equations. Note that the limit lines of Figure 1 are presented with core powers ranging from 80% of nominal up to 120% of nominal, although departure from nucleate boiling (DNB) is not the limiting condition tnroughout this range. Depending on the pressure, there is a power level below which hot leg saturation becomes the limiting condition, and at some paint, the main steam safety valves will actuate to preclude additional primary-side heatup. The OTAT reactor protection function is designed to prevent the occurrence of saturated conditions within the hot legs, and to preclude DNB conditions within the core for power levels not protected by the OPAT trip. The OPAT reactor protection function prevents operation above 118% of rated thermal power (RTP).

The " Adjusted" core thermal limits of Figure I were developed from the " Actual" limits and account for variations in the core axial power shape. During the development of the OTAT setpoint equation for South Texas, the value for the K1 gain (see WCAP-8745-P-A) was reduced to allow widening the f(edeadband. The deadband was widened to permit normal j plant operation without operating in the f(e penalty region. The DNB core thermal limits j were reduced to address the effects of the wider range of power shapes that can occur with the expanded fa dead-band. These reduced limits still represent the limiting DNBR of 1.38.

The partial derivatives used in the RETRAN DNB ratio (DNBR) model are based on the

" Adjusted" core limits and it has been verified that the OTAT setpoint equation protects these " Adjusted" core limits.

Figure 2 shows a comparison of the " Adjusted" core limits presented in Figure I and the conditions that correspond to the DNBR limit (1.38) as predicted by the RETRAN DNBR model. This model used the partial derivative coefficients presented in Table 1. Note that the plots of Figure 2 are presented in terms of reactor coolant system T VOA as a funClion of core power rather than core inlet temperature as a function of core power. The RETRAN ST!: 30897184

NOC-AB-CC0540 May 20,1999 Page 4 of 12 DNBR modelis applicable at or above full power conditions for all pressures between the low pressurizer pressure reactor trip setpoint (1845 psia) and the high pressurizer pressure reactor trip setpoint (2435 psia) assumed in the safety analysis. This comparison shows that the RETRAN DNBR model closely predicts the limiting DNBR at pressures equal to or greater than the nominal (2250 psia) pressure. At pressures below nominal pressure, the RETRAN DNBR model conservatively predicts the limiting DNBR to occur at temperatures that are less than the core limit values. With this, it is concluded that the partial derivative coefficients used in the RETRAN DNBR model are conservative for DNBR evaluations of applicable transients postulated for STP.

Table 1 RETRAN DNBR Calculation inputs N703849 l

l 703851 703852 l

Word 5 (CGAIN)

Word 5 (CGAIN) _l l -0.0524 0.005631 l

l l Word 6 (CPI) l 0.07428 l l 703853 l Word 5 (CGAIN) l 122.0 l l 703854 l 703858 l Word 6 (CPI) [ -619.27 l l Word 5 (CGAIN) l -0.027674 .l l 703897 l ,j Word 6 (CPI) 1.38 l STI: 30897184

NOC-AE-000540

' May 20,1999 Page 5 of 12 Figure 1 South Texas DNB Limit Lines (Actual vs. Adjusted) 640

. P = 2450 psia 630 -

g P = 2400 pela [A \

620-N*

• P = 2250 pela <Y % g s "

t, ,

610- g% l P = 2000 psia qY g s 4  % N N*N'Ni e

P = 1775 psia GY\

. s. *%,

\ *

• N h4 l 4

4  %  % \  %%

S 5' ' N g N =

Ng

• 580 -

N *

\

• g \\%

s

\ b * "k

% 4 \

670 - \ g h g  %

h \ \

660 - *

~

Adjusted g \

550 Adual

\

• q l

640-

\  %

l 530 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 Power (fraction of nominal)

STI: 30897184

NOC-AE-000540 May 20,1999 l Page 6 of 12 )

Figure 2 South Texas Core Umits (x,*8teted vs. RETRAN") ,

655 RETRAN P = 2450 pala*-'%, " "

P = 2400 pala* ,% Adjusted g,  %,,

t P = 2250 pala*

% =

.'*** re #,,  %,

F* .

'w I s,%,

615- 1 l { %g

(; P = 2000 psla*

-e f l l

w,'*-

, j,, ,

' )

b 595 - P = 1830 pela*

I I

, 'i ' =,,

• X i  %

==.. .

is i ,

I

" ", * * = d: e m

,7%

h

.* =

t .

• 575-I

%==*- m,* = '

I I I 4

: =* e i . . I I

1 i  :  :

I 1 1 555 ' ' ' ' '

O.8 0.85 0.9 0.95 1 1.05 1.1 1.15 1.2 Power (Fraction of Nominal) var P e 1830,2000, and 2250 paw the - --- Imes iderely the power traction below whch the Ernamg condson transfers from DNB to hot leg saturden, For P = 2400 and 2450 peu, the - - - anos identfy the power tracten below whch the Man Stearn Safsty Valves actude.

"'The RETRAN DNBA niodel coefficierts are appleable d power levels aquel to or greater than 100% of nommaL 4c. For MSLB:

Discuss how skewedflux shapes were evaluated.

For the HZP MSLB, the statepoint condition, with cold, non-uniform inlet temperatures and ARI, minus the worst stuck rod, is modeled in 3D each cycle. Similarly, for the full-power j MSLB statepoint condition, a 3D analysis is performed each cycle. Therefore, skewed flux shapes are explicitly modeled on a cycle-specific basis.

Discuss how initial core fuel and cladding temperatures were determined for input to RETRAN.

The revised thermal safety model described in WCAP-10851-P-A (Ref. 7) was used to calculate fuel and cladding temperatures using the PAD fuel performance code. This is the same method applied in evaluating fuel temperatures for recent South Texas Project core reload designs.

STI: 30897184

NOC-AE-000540 May 20,1999 Page 7 of 12 What minimum DNBR was determinedfor the MSLB andfor what break size, power level, and singlefailure assumptions.

Hot Zero Power Conditions Due to the asymmetric nature of this event, core power distribution has a significant impact on DNBR. A minimum DNBR was not explicitly calculated for the RETRAN hot zero power MSLB event. However it has been shown that the RETRAN MSLB (double-ended rupture) analysis results are significantly less limiting than the LOFTRAN i MSLB analysis results. Technical reasons for this are discussed in WCAP-14882 (Ref. 4). The peak heat flux fraction calculated in the RETRAN analysis was 5.4%.

compared to 11.8% calculated in the LOFfRAN analysis. The limiting single failure assumed in these analyses is the loss of one train of safety injection.

Full Power Conditions A spectrum of break sizes representing line and header breaks was examined in the full power MSLB analysis. The limiting break size was determined to be 1.09 ft 2 . The 4

DNBR results are 1.629 for the thimble cell and 1.666 for the typical cell. The corresponding DNBR Safety Analysis Limits are 1.38 (thimble) and 1.43 (typical), and the limiting single failure assumed is failure of one protection train.

t 4d.

If continuous operation with a feedwater heater out of service is planned, was the single failure of a secondfeedwater heater evaluated as part of the DNBR evaluations? What would be the effect on DNBRfrom failure of the secondfeedwater heater?

The single failure of a second feedwater heater was not explicitly examined. To account for the rare occasion when a high pressure feedwater heater is not available, a feedwater temperature range of 390 F to 440 F was covered in the analyses that support the A94 RSG Program. In the event of a second feedwater heater should fail, resulting in a feedwater temperature of approximately 378 F, sufficient DNB margin would continue to exist.

5.

Page 16 provides a table of minimum measuredflow and thermal design flow including the expected core bypassfor the two steam generator types. Describe how theflow rates

\

are determined. '

First a background discussion is provided defining the terms Best Estimate Flow (BEF).

Thermal Design Flow (TDF), and Minimum Measured Flow (MMF). A discussion of Core  !

Bypass methodologyis then presented. This is followed by a discussion of the differences between core bypass flows presented in the table. Finally, the specific answer to the question will be provided.

1 STI:30897184

[ ,

NOC-AE-000540 May 20,1999 Page 8 of 12 BACKGROUND DISCUSSION BEF is calculated by a Westinghouse computer model of the Reactor Coolant System (RCS) that includes STP plant-specific design parameters. These parameters account for vessel and fuel hydraulic design characteristics, system losses, RC pump hydraulic performance, and the hydraulic characteristics of the RCS. BEF represents an expected RCS flow value in a normally functioning plant at full power and nominal design conditions, and can be used as a point of reference for establishing minimum design flow values. The A94 replacement steam generators have more steam generator tubes than the Model E that they replace, which lowers the pressure drop across the primary side. This results in an increased BEF.

TDF is an assumed RCS flow value. It is chosen to be greater than the mmimum required )

flow to prevent DNB, but less than the minimum expected flow when considering conditions that might diminish RCS flow during plant life (increases in steam generator plugging levels, pump impeller smoothing, anticipated fuel changes, etc). The TDF is used in the safety analysis where a low RCS flow is conservative.

MMF is TDF increased by flow measurement uncertainty (MMF = TDF*[1+ uncertainty]). l The plant uses MMF as the minimum RCS flow Technical Specification (TS) surveillance value.

CORE BYPASS METHODOLOGY' Calculation of core bypass flow is performed to determine the amount of flow through each flow path not considered effective for cooling of the nuclear fuel. These flow paths include:

1. Flow through spray nozzles into upper head for head cooling purposes.

2. Flow through fuel assembly rod cluster control guide thimbles for cooling control rods.

3. Leakage flow from the vessel inlet nozzle, directly to the vessel outlet nozzle, through the gap between the vessel and the barrel.

4. Baffle and core barrel cooling flow that is not considered available for core cooling.

5. Flow in gaps between fuel assemblies on the core periphery and the adjacent baffle wall.

The Revised Thermal Design Procedure (RTDP) method uses the nominal manufacturing values in the calculation of this bypass flow. The Standard Design Procedure (STDP) uses the nominal manufacturing values with uncertainties in the conservative direction. 'Ihis accounts for the different core bypass values for the same model of steam generator when the different methodologies are used.

l DIFFERENCES BETWEEN CORE BYPASS FLOWS Mode E and A94 steam generators core bypass differences shown on the table are:

Increased flow to the upper head: To mitigate the impact of temperature on INCONEL

' 600 welds in the reactor vessel upper head, STP elected to reduce the upper head STI: 30897184

• NOC-AE-CD0340 May 20,1999 Page 9 of 12 temperature to the temperature of RCS cold leg. This required an increase in core bypass flow to the upper head.

Remove the rod cluster control guide thimble plugs from the fuel: The rod cluster control guide thimble plugs are being removed to reduced uplift forces on the fuel due to the higher RCS flow rate. This eliminates the need for stronger fuel assembly hold-down springs and reduces the potential for rod bowing experienced at STP.

ANSWER TO THE QUESTION To reduce the cost of replacement steam generators, STP elected to choose a TDF that would not require a re-analysis of the DNB events. To accomplish this goal, a TDF of 392,000 GPM was selected. This value ensures that the limiting DNB events which are modeled with the RTDP methodology would remain bounding because of a 0,4% increase in reactor core flow. DNB events modeled with STDP were shown to have sufficient margin to accommodate a 1.6% decrease in reactor core flow.

Based on the TDF determined above, the MMF was selected based on 2.8% measurement uncertainty. The MMF value is adequate to ensure that MMF can be maintained as steam generator tubes are plugged and anticipated plant conditions occur.

6. Of the non LOCA transients and accidents that were reanalyzed in NOC-AE-0080, for which event was the minimum DNBR calculated? What was the change in DNBR from \

this eventfor the steam generator replacement and how does the new value compare to the DNBR limitsforSouth Texas Project?

A summary of the minunum DNBR results for each DNB-limiting event reanalyzed is provided below.

Limiting Full Power Feedwater Malfunction (with Manual Rod Control):  ;

MDNBR(Model A94) = 1.604 (RETRAN-predicted)

MDNBR(Model E) = 1.777 (LOFFRAN-predicted)  !

MDNBR(Limit) = 1.38 Turbine Trip with Minimum Reactivity Feedback and Automatic Pressurizer Pressure Control:

MDNBR(Model A94) = 1.559 (RETRAN-predicted)

MDNBR(Model E) = 1.670 (LOFTRAN-predicted)

MDNBR(Limit) = 1.38 2

IIot Zero Power MSLB (1.4 ft double-ended rupture) at End of Cycle Conditions:

MDNBR(Model A94) = Not calculated (bounded by Model E)

MDNBR(Model E) =

2.43 - Thimble Cell (THINC-calculated)*

= 2.70 - Typical Cell (THINC-calculated)*

MDNBR(Limit) = 1.45 (w/o flow anomaly penalty)

STI: 30897184

• NOC-AE-0CD540 May 20,1999 Page 10 of 12 2 2 Full Power MSLB (1.09 ft for Model A94 and 1.06 ft for Model E) at End of Cycle Conditions:

MDNBR(Model A94) =

1.629 - Thimble Cell (THINC calculated)

= 1.666 - Typical Cell (THINC-calculated)

MDNBR(Model E) = 1.85 -

Thimble Cell (THINC-calculated)* l

= 1.89 -

Typical Cell (THINC-calculated)*

MDNBR(Limit) = 1.38 (Thimble Cell)

= 1.43 (Typical Cell)

• Cycle dependent; indicated values are for STP Unit 1 Cycle 9.

i For which event was the maximum reactor coolant system pressure calculated? What was the change in calculated reactor coolant system pressure for the steam generator replacement and how does the new pressure compare to the maximum allowable reactor coolant system pressurefor South Texas Project?

The turbine trip event with minimum reactivity feedback and no automatic pressurizer pressure control was most limiting with respect to the peak reactor coolant systera (RCS) pressure. The maximum calculated RCS pressure with the Model A94 steam generators (RETRAN) was 2743.5 psia. This pressure is 1.5 psiless than the maximum calculated RCS pressure with the Model E steam generators (LOFFRAN) and 5 psi less than the maximum allowable RCS pressure.

7.

Address each of the three conditions listed in the conclusions section of the NRC staff SERfor WCAP-14882.

The NRC staff concludes in the SERfor WCAP-14882 (Ref. 5) that the "use ofRETRAN as described in WCAP-14882 is acceptablefor licensing calculations and RETRAN may be used to replace the LOFTRAN computer code in Westinghouse reload methodology provided that thefollowing conditions are met:

1. The transients and accidents that Westinghouse proposes to analyze with RETRAN are listed in this SER (Table 1) and the NRC staff review of RETRAN usage by Westinghouse was limited to this set. Use of this codefor other analyticalpurposes willrequire additionaljustification.

2.

WCAP-14882 describes modeling of Westinghouse designed 4., 3, and 2 loop plants of the type that are currently operating. Use of the code to analyze other designs, including the Westinghouse AP600, willrequire additionaljustification.

3 l Conservative safety analyses using RETRAN are dependent on the selection of \

conservative input. Acceptable methodology for developing plant-specific input is l discussed in WCAP-14882 and in Reference 6. Licensing applications using \

RETRAN should include the source of andjustificationfor the input data used in the i

analysis."

l sTI: 30897184 l

NOC-AE-000540

,- ~* . May 20,1999 Page 11 of 12 i

l Each of these conditions are addressed below, as they relate to the South Texas Project Model A94 Replacement Steam Generator Program.

1.

The non-LOCA transients explicitly analyzed with RETRAN for this program include the

{

following: feedwater system malfunctions, steam system piping failures, turbine trip, loss I of offsite power, loss of normal feedwater flow, and feedwater system pipe break. All of these events are listed in Table 1 of the SER; therefore, no additional justification is required.

2.

The South Texas Project nuclear units are 4-loop, Westinghouse-designed, pressurized water reactors that are currently in commercial operation. Therefore, no additional justification is required.

3. The non-LOCA RETRAN analyses were performed in accordance with the methodologies discussed in WCAP-14882 (Ref. 4) and WCAP-9272-P-A (Ref. 6).

I 1

I I

l STI: 30897184 )

• ' NOC-AE-000540 May 20,1999 Page 12 of 12 References 1.

WCAP-8745 P-A, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," Ellenberger, S. L., et al., September 1986.

2.

WCAP-7907-P-A, "LOFTRAN Code Description," Burnett, T. W., et al., April 1984.

3.

Letter NSD-NRC-98-5765, H. A. Sepp, (Westinghouse) to T. E. Collins (NRC), " Responses to Request for AdditionalInformation on WCAP-14882, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses" [ Proprietary], August 26,1998.

4.

WCAP-14882 Revision 0, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," Huegel, D. S., et al., June 1997.

5.

USNRC Letter, " Acceptance for Referencing of Licensing Topical Report WCAP-14882,

'RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis' (TAC NO. M99107)," Akstulewicz, F. (USNRC) to Sepp, H. (W),

February 11,1999.

6.

WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology," Bordelon, F. M., et

. al., Approved July 1985.

7.

WCAP-10851-P-A, " Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," Weiner, R. A., et al., Approved August 1988.

l STI: 30897184