ML071060392

Request for Approval of Topical Report DOM-NAF-5, Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (Kps.)

ML071060392
Person / Time
Site:	Kewaunee  (DPR-043)
Issue date:	04/16/2007
From:	Gerald Bichof Dominion, Dominion Energy Kewaunee
To:	Document Control Desk, NRC/NRR/ADRO
References
06-578B DOM-NAF-5
Download: ML071060392 (52)

Text

1)ominion Energy Kewaunee, Inc.

i 0 0 0 I ) o n i ~ r i ~ o n Boulevard. Glcn Allcn, VA LiO6o A p r i l 16, 2007 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 1

Dominion' Serial No.

06-578B NL&OS/CDS: RO Docket No.

50-305 License No. DPR-43 DOMINION ENERGY KEWAUNEE. INC.

KEWAUNEE POWER STATION REQUEST FOR APPROVAL OF TOPICAL REPORT DOM-NAF-5. "APPLICATION OF DOMINION NUCLEAR CORE DESIGN AND SAFETY ANALYSIS METHODS TO THE KEWAUNEE POWER STATION (KPS)"

In a January 31, 2006, public meeting with NRC staff, Dominion Energy Kewaunee (DEK) presented a conceptual approach and implementation strategy for application of approved nuclear core design and safety analysis methods to Kewaunee Power Station (KPS) (reference 1).

Fundamental to the proposed approach was creation of a composite topical report (DOM-NAF-5) that would document the application of the relevant methodologies to KPS.

On August 16, 2006, DEK submitted DOM-NAF-5 without attachments A and B (reference 2). On December 6, 2006, Attachment A to DOM-NAF-5, containing CMS benchmark analysis results, was submitted (reference 3).

Attachment B to DOM-NAF-5, containing RETRAN benchmark analysis results, is completed and is attached. This submittal, in conjunction with References 2 and 3, provides the complete contents of topical report DOM-NAF-5.

DEK will issue a consolidated version of the topical report, DOM-NAF-5, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)," including the supplemental material of Attachments A and B, following NRC review and approval.

DEK is requesting approval of DOM-NAF-5 with the intent of subsequently implementing the Dominion Statistical DNBR Evaluation Methodology with VIPRE-D at KPS. The plant specific application of this methodology will be submitted separately and will employ the VIPRE-D code with the Westinghouse WRB-1 Critical Heat Flux correlation for the thermal-hydraulic analysis of Westinghouse 422V+ fuel assemblies at KPS.

In order to support application of these methods to KPS Cycle 29, DEK requests NRC staff review and approval of DOM-NAF-5 by September 30, 2007.

A subsequent administrative license amendment request (LAR) to include DOM-NAF-5 among the reference methodology reports in the KPS Technical Specifications is scheduled to be submitted in June 2007. The requested date for NRC staff approval of that LAR will be January 31, 2008. The LAR approval date supports application of DOM-NAF-5 to KPS Cycle 29, which is scheduled to begin in April 2008.

Serial No. 06-5788 Request for Approval of DOM-NAF-5 Page 2 of 2 Should you have any questions, please contact Mr. Craig D. Sly at 804-273-2784.

Very truly yours, G. T. Bischof u

Vice President - Nuclear Engineering

References:

1. Summary of Meeting on January 31, 2006, "To Discuss the Applicability of Dominion Safety and Core Design Methods to Kewaunee Power Station," (TAC No. MC 9566),

(ADAMS Accession Number ML060400098).

2. Letter from G. T. Bischof (DEK) to NRC, "Request for Approval of Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated August 16, 2006 (ADAMS Accession Number ML062370351).

3. Letter from G. T. Bischof (DEK) to NRC, "Attachment A to Topical Report DOM-NAF-5, 'Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS),"' dated December 6, 2006 (ADAMS Accession Number ML0063410177).

Attachment:

1. DOM-NAF-5, Attachment B, "RETRAN Benchmarking Information," dated March 2007.

Commitments made in this letter: None cc:

Regional Administrator U. S. Nuclear Regulatory Commission Region Ill 2443 Warrenville Road Suite 21 0 Lisle, Illinois 60532-4352 Mr. R. F. Kuntz Project Manager U.S. Nuclear Regulatory Commission Mail Stop 0-7D1 A Washington, D. C. 20555 Mr. S. C. Burton NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 06-0578B TOPICAL REPORT DOM-NAF-5 ATTACHMENT B RETRAN BENCHMARKING INFORMATION DOM-NAF-5, APPLICATION OF DOMINION NUCLEAR CORE DESIGN AND SAFETY ANALYSIS METHODS TO THE KEWAUNEE POWER STATION (KPS),

DATED MARCH 2007 KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Attachment B RETRAN Benchmarking Information Prepared by:

John C. Lautzenheiser Mark C. Handrick Approved:

~uiervisor, Nudear Safety Analysis

DOM-NAF-5 Attachment B RETRAN Benchmarlung Information Table of Contents TABLE OF CONTENTS..................................................................................................... 2 1.0 Introduction and Summary 3

1.1 Introduction 3

1.2 Summary 3

2.0 KPS RETRAN Model................................................................................................... 4 3.0 Method of Analysis..................................................................................................... 8 4.0 Demonstration Analysis Results................................................................................... 9 4.1 Loss of Load 9

4.2 Locked Rotor..........................................................................................................

15 4.3 Loss of Normal Feedwater.......................................................................................

24 4.4 Main Steam Line Break 30 4.5 Control Rod Bank Withdrawal at Power..................................................................

37 4.6 Loss of Flow........................................................................................................... 44 5.0 Conclusions 49 6.0 References 49

DOM-NAF-5 Attachment B RETRAN Benchmarking Information 1.0 Introduction and Summary 1.1 Introduction Topical report VEP-FRD-4 1, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code," (Reference 1) details the Dominion methodology for Nuclear Steam Supply System (NSSS) non-LOCA transient analyses. This methodology encompasses the non-LOCA licensing analyses required for the Condition I, 11, 111, and IV transients and accidents addressed in the Updated Safety Analysis Report (USAR). The VEP-FRD-41 methods are also used in support of reload core analysis. In addition, this capability is used to perform best-estimate analyses for plant operational support applications. The material herein supplements the applicability assessment of RETRAN methods for Kewaunee Power Station (KPS) that is presented in Section 3.4 of DOM-NAF-5, demonstrating that the VEP-FRD-41 methods are acceptable for the stated applications.

1.2 Summary This report provides a description of the RETRAN base model for KPS and results of demonstration analyses using this model. The KPS model was developed in accordance with the methods in VEP-FRD-41, with certain noding changes noted below. This assessment reaffirms the conclusion in Section 3.4 of DOM-NAF-5, that the Dominion RETRAN methods, as documented in topical report VEP-FRD-41, are applicable to KPS and can be applied to KPS licensing analysis for reload core design and safety analysis. Dominion analyses of KPS will employ the modeling in VEP-FRD-41, as augmented with the noding changes listed below.

Thus, VEP-FRD-41, as augmented, is the Dominion methodology for analyses of non-LOCA NSSS transients for KPS.

The KPS RETRAN base model contains the following alterations in noding with respect to the modeling that is documented in VEP-FRD-41.

a) The KPS model explicitly models the safety injection (SI) accumulators.

b) The KPS model has separate volumes for the steam generator inlet and outlet plenums.

C) The KPS model includes cooling paths between downcomer and upper head (Main Steam Line Break overlay).

DOM-NAF-5 Attachment B RETRAN Benchmarking Information 2.0 KPS RETRAN Model A KPS RETRAN-02 Base Model and associated model overlays are developed using Dominion analysis methods described in the Dominion RETRAN topical report (Reference I). The Dominion analysis methods are applied consistent with the conditions and limitations described in the Dominion topical report and in the applicable NRC Safety Evaluation Reports (SERs).

The KPS Base Model noding diagram is shown on Figure 2-1. Volume numbers are circled, junctions are represented by arrows, and the heat conductors are shaded. This model simulates both reactor coolant system (RCS) loops and has a single-node steam generator (SG) secondary side, consistent with Dominion methodology. The SG primary nodalization includes 10 steam generator tube volumes and conductors. There is a multi-node SG secondary overlay that can be added to the Base Model for sensitivity studies although none of the analysis results presented herein utilize this overlay.

In addition to the base KPS model, an overlay deck is used to create a split reactor vessel model to use when analyzing Main Steam Line Break (MSLB) events, consistent with Dominion methodology. This overlay adds volumes to create a second, parallel flow path through the active core from the lower plenum to the upper plenum such that RCS loop temperature asymmetries can be represented. This overlay also includes flow paths between the downcomer and the upper head to model the small amount of cooling flow to the upper head. These flowpaths may also be added for other events when flashing in the upper head is expected to occur. A noding diagram of the split reactor vessel is shown on Figure 2-2. This figure shows the hot leg volumes (101, 201) and cold leg volumes (1 16, 216) so the reactor vessel can be seen in context of the RCS interface. Otherwise, the noding for all other regions of the model are unchanged from Figure 2-1.

The base KPS model noding is virtually identical to the Surry (SPS) and North Anna (NAPS) models with the exception of some minor noding differences listed as follows, which are updated from the original list provided in DOM-NAF-5 Section 3.4.1.4.

a) The KPS model explicitly models the SI accumulators.

b) The KPS model has separate volumes for the SG inlet and outlet plenums.

c) The KPS model includes cooling paths between downcomer and upper head (MSLB overlay).

The SI accumulators are part of the KPS model because injection from the accumulators is more likely to occur during a MSLB cooldown event for a two-loop plant. The cooling paths are included in the MSLB overlay to appropriately model the effects of flashing in the head, as noted above. The use of separate volumes for the inlet and outlet should have little effect on transient

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg 5 of 49 response since the fluid temperature in these volumes is generally the same as the connecting RCS piping.

The Dominion models, including the KPS model, have some differences compared to the vendor RETRAN model that was used to perform the current USAR analyses. Table 2-1 and the subsequent text discussion provides an overview of these differences. Additional details concerning differences between the Dominion KPS and USAR RETRAN models are discussed in the demonstration analyses in Section 4.

A description of the Dominion RETRAN methodology is provided in Reference 1, where specific model details are discussed in Sections 4 and 5 of that reference.

Table 2-1 Dominion USAR RETRAN Model Com~arison Parameter I Dominion 1 USAR Noding:

Reactor Vessel active core (special split core overlay for MSLB only)

Single flow path - 3 axial nodes for axial nod&

for active core.

Increased nodalization in other Steam Generator I

I Heat Transfer Option 1 Forced 1 Forced + Free Convection HT Split (two parallel flow path) - 4 Reactivity Model Doppler Feedback Moderator Feedback Decay Heat Other:

Single node secondary. Five axial levels (10 nodes) for SG tubes vessel regions.

Multi-node secondary. Four axial levels for SG tubes (primary and primary side. Local Conditions Heat Transfer model available for loss of heat sink events.

Doppler temperature coefficient that is a function of T F " ~ ~.

Moderator temperature coefficient ANS-5.1 1979 Standard U-235 with 1500 day bum.

Q = 190 MeVIfission.

Bounds additional 20 uncertainty Gap Expansion Model secondary).

Doppler-only power coefficient and a

Doppler temperature coefficient effect driven by moderator temperature.

Moderator density coefficient ANS-5.1 1979 Standard Equilibrium decay heat, Bounds additional 20. uncertainty Used for HZP events only.

Map Used for all events.

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 7 of 49 Figure 2-2. KPS Split Vessel Nodalization

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 8 of 49 3.0 Method of Analysis As discussed in Section 3.4.3 of DOM-NAF-5, validation of the Dominion KPS RETRAN method involves comparison of RETRAN analyses to the KPS USAR analysis of record (AOR) for select events. These events represent a broad variation in behavior (e.g. RCS heatup, RCS cooldown/depressurization, reactivity excursion, loss of heat sink, etc.), and demonstrate the ability to appropriately model key phenomena for a range of transient responses. The transients selected for comparison with their corresponding KPS USAR section are provided in Table 3-1. For each transient, an analysis is performed using the Dominion KPS RETRAN model and analysis methods. Initial conditions are established to be consistent with the input used in USAR analyses.

Table 3-1 Transients Analyzed for USAR Comparison Transient 1 KPS USAH Section I Control Rod Withdrawal at Power 1 14.1.2 I

I Loss of Flow 1 14.1.8 1

I Locked Rotor 1 14.1.8 I

I Loss of LoadTurbine Trip 1 14.1.9 1 Loss of Normal Feedwater 1 14.1.10 1

I Main Steam Line Break 1 14.2.5 I

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 9 of 49 4.0 Demonstration Analysis Results A summary for each transient comparison is presented in the following sections. Included in each section is an input summary identifying key inputs and assumptions along with differences from USAR assumptions. A comparison of the results for key parameters is provided with an explanation of key differences between the Dominion and USAR cases.

4.1 Loss of Load The Loss of LoadtTurbine Trip (LOL) event is defined as a complete loss-of-steam load and turbine trip from full power without a direct reactor trip, resulting in a primary fluid temperature rise and a corresponding pressure increase in the primary system. This transient results in degraded steam generator heat transfer, reactor coolant heatup and pressure increase following a manual turbine trip.

The LOL transient scenario presented here was developed to analyze primary RCS overpressurization. It is initiated by decreasing both the steam flow and feedwater flow to zero immediately after a manual turbine trip. The input summary is provided in Table 4.1-1.

Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.1-1 LOL I n ~ u t Summarv 1 Initial Conditions I

I 1

Parameter I (delays trip)

Value NSSS Power (MW)

RCS Flow (gprnlloop)

Vessel T (F)

RCS Pressure (psia)

Notes 1815.6 89,000 579 2200 Pressurizer Level (%)

SG Level (%)

SG Pressure (psia)

Assumptions1Configuration Reactor trip Automatic rod control Pressurizer sprays, PORVs Main steam dumps, SG PORV AFW flow Includes 2% uncertainty and pump heat Thermal Design Includes +6 F uncertainty Includes -50 psia uncertainty SG tube plugging (%)

53 44 836.8 10

( Max value Reactivity Parameters Doppler Temp. Coefficient (pcm/F)

Includes +5 % uncertainty USAR = 797.98 only Hi Pzr Pressure is active Not credited Not credited Not credited Not credited Moderator Temp. Coefficient (pcm/F)

-2.5, Most negative USAR uses least negative Doppler-only power coefficient, and a least negative 0.0 DTC (driven by moderator temp).

USAR uses 0.0 Ak/(gmlcc) for MDC

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 10 of 49 Results - LOL Pressure in the RCS increases during a LOL due to degraded heat transfer in the steam generator and is alleviated only when the pressurizer safety valves (PSV) open. The pressurizer pressure response is shown on Figure 4.1-1 and the peak RCS pressure values are listed in Table 4.1-1. Pressure for the Dominion case increases slightly faster initially but peaks at about the same time and same value as the USAR data.

Figure 4.1-2 shows the power response which is nearly constant until a reactor trip on high pressurizer pressure occurs. The Dominion case trips slightly earlier than the USAR data because of the higher RCS pressurization rate.

The core average temperature is shown on Figures 4.1-3. The Dominion and USAR temperature are virtually identical until after the reactor trip and peak pressure occurs, at which time they diverge somewhat but trend together. Because of the RCS heat up and coolant expansion, there is a liquid insurge to the pressurizer as shown by the pressurizer liquid volume increase on Figure 4.1-4. The pressurizer liquid volume increases faster for the Dominion case early in the event, yielding the slightly faster pressure increase discussed earlier. By the time the peak pressure is reached at approximately 11 seconds, the liquid volumes for both cases compare well. The difference after that time is consistent with the temperature response.

The steam generator pressure is shown on Figure 4.1-5. The steam generator heat transfer degradation is strongly related to the secondary pressure increase (saturation temperature increase) since the dominant secondary heat transfer mode is boiling heat transfer. The Dominion case starts from a higher initial pressure, which is the result of model initialization to match the design heat transfer surface area for the single node steam generator, and peaks at a higher pressure than the USAR case.

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Table 4.1-2 LOL Primary Overpressure Results I

Parameter I

DOM I

USAR 1

Sequence of Events:

Reactor Trip (sec) pg. 11 of 49 (High Pzr Pres)

Peak RCS Pressure (psia)

Peak MSS Pressure (~sia)

Summary - LOL 8.4 The Dominion Kewaunee analysis provides results that are similar to the USAR analysis for the LOL event. The RCS temperatures agree well early in the event and although they diverge somewhat later in the event, this does not occur until after the RCS pressure peak occurs and pressure relief begins. There are small differences in pressurization rates early in the event; however, the peak RCS pressure values are the same for both cases. In addition, the peak SG pressure is slightly higher for the Dominion case. There is adequate margin to the RCS pressure acceptance criterion of 2750 psia.

8.9 2697 1192 2697 1182

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 12of49 Figure 4.1-1 LOL - Pressurizer Pressure 2700 2100 J 1

0 2

4 6

8 10 12 14 16 18 20 Time (seconds)

Figure 4.1 -2 LOL - Reactor Power 2

4 6

8 10 12 14 16 18 20 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 13of49 Figure 4.1-3 LOL - RCS Average Temperature 570 0

2 4

6 8

10 12 14 16 18 20 Time (seconds)

Figure 4.1 -4 LOL - Pressurizer Liquid Volume 0

2 4

6 8

10 12 14 16 18 20 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.1-5 LOL - Steam Generator A Pressure pg. 14 of49 700 -1 0

2 4

6 8

10 12 14 16 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 15 of49 4.2 Locked Rotor The Locked Rotor 1 Shaft Break (LR) event is defined as an instantaneous seizure of a Reactor Coolant Pump (RCP) rotor, rapidly reducing flow in the affected reactor coolant loop leading to a reactor trip on a low-flow signal from the Reactor Protection System. The event creates a rapid expansion of the reactor coolant and reduced heat transfer in the steam generators, causing an insurge to the pressurizer and pressure increase throughout the reactor coolant system (RCS).

The LR transient scenario presented here was developed to analyze primary RCS overpressurization.

It is initiated by setting one RCP speed to zero as the system is operating at full power. The reactor coolant low loop flow reactor trip is credited, with a setpoint of 86.5% of the initial flow. The input summary is provided in Table 4.2-1. Most of the input parameters are the same as those used in the USAR Chapter 14 analyses. Where differences from the USAR inputs exist, they are indicated in the Notes column.

Table 4.2-1 LR I n ~ u t Summarv Parameter I Value 1 Notes Initial Conditions NSSS Power (MW)

RCS Flow (gprnlloop)

Vessel TAVC (F)

Initial Fuel Average Temperature (F)

RCS Pressure (psia)

Pressurizer 1,evel (%70'1 SG Level (%)

SG Pressure (psia)

Assumptions/Configuration Reactor trip Automatic rod control Pressurizer sprays, PORVs Main steam dumps, SG PORV AFW flow SG tube plugging (%)

RCPImotor moment of inertia (lbm/ftz)

Reactivity Parameters Doppler Temp. Coefficient (pcm/F)

Moderator Temp. Coefficient (pcm/F) 1815.6 89,000 579 1332 2300 48 Includes pump heat and 2% uncertainty Thermal Design Flow Includes +6 F uncertainty Includes +50 psia uncertainty Nominal 1 - SG pressure adjusted for minimal change to the SG tube surface areas by steady-state initialization 44 836.8 (Note 1) 10 72,000

-1.2, Least Negative 0.0 Nominal Only Low RCS Loop Flow is credited Not credited Not credited Not credited Not credited Max value 90% of nominal USAR uses most negative Doppler-only power coefficient, and a least negative DTC (driven by moderator temp).

USAR uses 0.0 Akl(gm/cc) for Moderator Density Coefficient

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 16of49 Results - LR RCS Overpressure Case Pressure in the RCS increases during a LR event due to degraded heat transfer in the steam generator and is alleviated only when the pressurizer safety valves (PSV) open. The pressurizer pressure response is shown on Figure 4.2-1 and the peak RCS pressure values are listed in Table 4.2-1. Pressure for the Dominion case increases slightly faster initially but peaks at about the same time and approximately the same value as the USAR data.

Figure 4.2-2 shows the pressure response in the reactor vessel lower plenum, which compares well to the USAR data, particularly until the point of peak pressure.

Figures 4.2-3 and 4.2-4 show the total core inlet volumetric flow and faulted loop volumetric flow (fraction of nominal), respectively. The predicted volumetric core inlet volumetric flow rate decreases more rapidly than the USAR data. This behavior is also present for the faulted loop volumetric flow, where the loop flow reverses earlier than the USAR data. This is conservative behavior in comparison to the USAR data. Each analysis assumes 90% of nominal RCP inertia to limit the coastdown.

Figure 4.2-5 shows the core thermal power response, which matches the USAR analysis well, except for small differences prior to the reactor trip. The USAR Doppler feedback model is a function of (1) Doppler-only power coefficient (DPC), and (2) Doppler temperature coefficient (DTC). The DTC modeled in the USAR analysis is actually a function of moderator temperature rather than fuel temperature. The Dominion reactivity model uses a Doppler Temperature Coefficient, dependent only on changes in fuel temperature, which provides the prompt feedback component. The Dominion DTC model is described in Section 5.13 of Reference 1. The Dominion model predicts core power to decrease prior to reactor trip, which is expected due to negative Doppler feedback as the fuel temperature increases.

The computed core average heat flux shown in Figure 4.2-6 compares well with the USAR data. The small difference in the core heat flux response during the first second of the transient is probably due to the difference in modeling of the core heat transfer coefficients (HTC). In the USAR analysis, the core HTCs are held fixed at their initial values. In the Dominion model, the forced convection HTCs are allowed to decrease with the decaying RCS flow rate, effectively reducing the core heat flux during the first second of the event.

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 17 of49 The faulted loop hot leg and cold leg temperatures are shown on Figure 4.2-7. The Dominion and USAR temperature are virtually identical until time when the peak pressure occurs, at which time they diverge somewhat but trend together.

A summary of the LR transient analysis comparison is provided in Table 4.2-2.

Table 4.2-2 LR RCS Overpressure Results I

Parameter DOM USAR 1

Sequence of Events:

\\

/ Peak RCS Pressure (psia) 268 1 2683 Reactor Trip on Low RCS Flow (sec)

Peak RCS Pressure (set)

Summary - LR RCS Overpressure Case The Dominion Kewaunee analysis provides results that are similar to the USAR analysis for the LR event. There are small differences in RCS coastdown flow and core power response during the early portion of the transient. However these differences occur prior to the time of peak RCS pressure (at approximately 4 seconds), and are relatively insignificant for this transient. The peak RCS pressure values are essentially the same for both cases. The predicted peak RCS pressure for the Dominion model (2681 psia) is just slightly below the USAR peak pressure (2683 psia). In each case, there is adequate margin to the RCS peak pressure acceptance criterion of 2750 psia.

0.80 4.2 0.80 4.5

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.2-1 LR - Pressurizer Pressure pg. 18of49 4

6 8

Time (seconds)

Figure 4.2-2 LR - Lower Plenum Pressure 2800 I 4

6 8

Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.2-3 LR - Core Inlet Volumetric Flow pg. 19of49 4

6 8

Time (seconds)

Figure 4.2-4 LR - Faulted Loop Volumetric Flow DOM USAR 4

6 8

Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.2-5 LR - Core Power pg. 20 of 49 4

6 8

Time (seconds)

F~gure 4.2-6 LR - Core Heat FLUX p

USAR 4

6 8

T~me (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.2-7 LR - Faulted Loop Temperatures pg. 21 of49 USAR HL Cold Leg 4

6 8

Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 22 of 49 LR Peak Cladding Temperature The Locked Rotor event is also analyzed to demonstrate that a coolable core geometry is maintained. Acceptance criteria for this analyses are met by showing that the peak cladding temperature (PCT) remains below 2700 O F, and that the oxidation level is below 16.0 percent by weight. A hot spot evaluation is performed to calculate the peak cladding temperature and oxidation level. The Dominion Hot Spot model is described in Topical Report VEP-NFE-%A, "VEPCO Evaluation of the Control Rod Ejection Transient."

(Reference 2) The Dominion Hot Spot model was used to evaluate the Kewaunee PCT and oxidation level for the LR event.

The Dominion hot spot model is used to predict the thermal-hydraulic response of the fuel for a hypothetical core hot spot during a transient. The hot spot model describes a one-foot segment of a single fuel rod assumed to be at the location of the peak core power location during a transient. The hot spot model uses boundary conditions from the LR system transient analysis to define inlet flow and core average power conditions. The hot spot model uses Kewaunee-specific values for fuel dimensions, fuel material properties, fluid volume, and junction flow areas.

The hot spot model is run to 0.001 seconds and a restart file is saved. Upon restart, the fuelkladding gap conductance (thermal conductivity) is modified to simulate gap closure by setting the gap heat transfer coefficient to 10,000 ~ t u / f t ~ - h r - " ~

for a gap conductance of 3.125 Btulft-hr-OF. The hot spot model input summary is provided in Table 4.2-3. Most of the input parameters are the same as those used in the USAR Chapter 14 analyses. Where differences from the USAR inputs exist, they are indicated in the Notes column.

Table 4.2-3 Hot S ~ o t Model I n ~ u t Summarv Parameter Computer Code Used Initial Conditions Ratio of Initial to Nominal Power RCS Thermal Design Flow (gprnlloop)

Hot Spot Peaking Factor Assumptions/Configuration Pre-DNB Film Heat Transfer Coefficient Value RETRAN-02 1.02 Time of DNB (sec)

Notes USAR uses FACTRAN 89,000 2.50 Thorn Post DNB Film Boiling Heat Transfer Coefficient Fuel Pin Model Post DNB Gap Heat Transfer Coefficient USAR uses Dittus-Boelter or 0.001 Bishop-Sandberg-Tong 10,000 Jens-Lottes

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 23 of 49 LR Peak Cladding Temperature Results (Btulhr-ft2-OF)

Gap Thermal Expansion Model activated?

Zircaloy-Water Reaction activated?

The peak cladding inner surface temperature obtained from the Dominion Kewaunee hot spot model is 1633 OF. The maximum zircaloy-water reaction depth into the cladding is 2.04E-06 feet, which corresponds to approximately 0.10% by weight based on the nominal cladding thickness of 2.025E-03 feet. A summary of the LR Peak Cladding Temperature Hot Spot analysis comparison is provided in Table 4.2-4.

Yes Yes Table 4.2-4 LR Hot S ~ o t Results The Dominion peak cladding temperature and maximum oxidation values are less than the USAR values, however both cases demonstrate considerable margin to the acceptance criterion of 2700 OF and 16.0% by weight, respectively.

Parameter Peak Cladding Temperature Maximum Zr-water reaction (w/o)

The difference in zirconium-water reaction results is understood by examination of the Baker-Just parabolic rate equation, which shows that the zirconium-water reaction becomes significant above cladding temperatures of 1800 OF. Since the Dominion hot spot model results do not predict peak cladding temperatures of this magnitude, it is expected that the corresponding zirconium-water reaction would be less than the US AR analysis. In the RETRAN-02 code, the rate of change in the reacting metal-oxide interface is proportional to exp(-41200lT) where T is the absolute clad temperature (OR).

The ratio of exp[-41200/(1633+460)] / exp[-4 1200/(1900+460)] = 0.1. This value corresponds approximately to the oxidation ratio of the Dominion vs. USAR result.

DOM 1633 OF 0.10 USAR 1900 O F 0.61

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 24 of 49 4.3 Loss of Normal Feedwater The Loss of Normal Feedwater (LONF) event causes a reduction in heat removal from the primary side to the secondary system. Following a reactor trip, heat transfer to the steam generators continues to degrade resulting in an increase in RCS fluid temperature and a corresponding insurge of fluid into the pressurizer. There is the possibility of RCS pressure exceeding allowable values or the pressurizer becoming filled and discharging water through the relief valves. The event is mitigated when Auxiliary Feedwater (AFW) flow is initiated and adequate primary to secondary side heat removal is restored. This analysis shows that the AFW system is able to remove core decay heat, pump heat and stored energy such that there is no loss of water from the RCS and pressure limits are not exceeded. The LONF input summary is provided in Table 4.3-1. Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.3-1 LONF I n ~ u t Summarv Parameter Initial Conditions NSSS Power (MW)

RCS Flow (gprnlloop)

Vessel T A " ~

(F)

RCS Pressure (psia)

Pressurizer Level (%I SG Pressure (~sia)

SG Level (%)

Low-Low Level Reactor Trip Setpoint Pressurizer: smavs. heaters. PORVs AFW Temperature (F)

AFW Pump configuration Auxiliary feedwater flow rate (gpm)

AFW Delay after Low-Low-SG level (sec)

Local Conditions Heat Transfer model Steam Generator MSSV blowdown (%)

Reactivity Parameters Doppler Temp. Coefficient (pcm/F)

Moderator Temp. Coefficient (pcm/F)

Value I Notes 1790.62 1 Includes 0.6% unc. and DumD heat 89.000 1 Thermal Design 0%

I Percent of narrow range span 579 2300 5 3 863.4 5 1 I assumed o~erable HFP nominal + 6 F Nominal + 50 psi Nominal + 5%

USAR= 829.0 Nominal + 7%

800 I Max delay active

( SG secondary side 120 max value one motor-driven pump per SG variable as function of SG mess.

negative only power coefficient, and a least negative DTC (driven by moderator tem~).

3 USAR= multi-node SG USAR= blowdown not modeled 0.0 USAR uses 0.0 Akl(grnlcc) for Moderator Density Coefficient

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 25 of 49 Results - LONF The results for the LOW comparison analysis are presented in Table 4.3-2 and Figures 4.3-1 through 4.3-6. The loss of feedwater flow to the steam generators (SG) results in a reduction in SG level until a reactor trip occurs on Low-Low SG level. Normalized power is shown on Figure 4.3-1. The power response is similar for the Dominion and USAR cases, except that the trip for the Dominion case occurs about 10 seconds later due to slightly higher initial fluid mass in the Dominion SG secondary side. (Higher initial mass has been demonstrated in a sensitivity case to yield conservative results due to delayed reactor trip).

The reduction in SG level results in degraded heat transfer from the primary to secondary systems and an increase in RCS temperature, plotted on Figure 4.3-2. The heatup prior to reactor trip is more pronounced for the Dominion model due in part to the delay in reactor trip. After the reactor trip occurs, the RCS cools somewhat until the loss of SG level and related heat removal is no longer able to remove decay and residual heat. The temperature then increases until AFW flow is actuated and adequate heat removal is restored.

The effect of the temperature change is reflected in the fluid density and associated pressurizer level change, as seen on Figure 4.3-3. The initial pressurizer insurge is higher for the Dominion case, which receives a later reactor trip signal as noted earlier. The maximum pressurizer level, which occurs after the reactor trip, is higher for the USAR case.

This appears to be due primarily to higher pressurizer spray flow rates and more conservative decay heat assumptions for the USAR cases. Note, both the Dominion and USAR methods use decay heat profiles that conservatively bound the values for the 1979 ANS-5.1 decay heat model plus Zsigma uncertainty; however, the USAR method uses a higher value for the decay heat multiplier. Also, the Dominion methodology assumes a conservative value for pressurizer spray flow, although the USAR model appears to apply additional conservatism.

Next, pressurizer pressure responds to the level insurge as shown on Figure 4.3-4. The initial pressure increase for the Dominion case is sufficiently high to cause the pressurizer Power Operated Relief Valves (PORV) to open. The subsequent Dominion pressure response is below the USAR profile but eventually increases above the USAR values during the second insurge. Overall, the pressure response for the USAR case is somewhat flatter due to the higher pressurizer spray flow, which tends to suppress the pressure increases.

Also, the delay in reactor trip causes the initial pressure increase for the Dominion case to be more severe.

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 26 of 49 The secondary side response for SG mass and pressure is shown on Figures 4.3-5 and 4.3-6, respectively. Other than the differences in initial pressure and fluid mass, the responses for the Dominion case and USAR case are similar. Note, the Dominion case models Main Steam Safety Valve. (MSSV) blowdown, resulting in slightly lower steady state SG pressures and a more clearly defined valve opening and closing response. Finally, as noted in the plot for SG mass, once AFW flow is initiated, the SG level gradually increases and adequate heat removal is eventually restored.

Table 4.3-2 LONF Results Parameter Se uence of Events:

Loss of Feedwater (sec)

Reactor Trip (sec)

(Low-Low SG Level)

Summary - LONF peak RCS Pressure (psis)

Peak PZP Liauid Volume (ft3)

The Dominion analysis provides results that are similar to the USAR analysis for the LONF event. Differences in the maximum pressurizer level are explained primarily by differences in pressurizer spray assumptions and decay heat modeling. Both analyses are conservative and demonstrate adequate margin to pressurizer overfill acceptance criteria.

2433 842 234 1 925

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.3-1 LONF - Core Power pg. 27 of49 100 Time (seconds)

Figure 4.3-2 LONF - RCS Average Temperature USAR DOM 100 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 28 of49 Figure 4.3-3 LONF - Pressurizer Liquid Volume 1200 100 Time (seconds)

Figure 4.3-4 LONF - Pressurizer Pressure 100 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 29 of49 Figure 4.3-5 LONF - Steam Generator Mass 120000 0 4 1

10 100 1000 Time (seconds)

Figure 4.3-6 LONF - Steam Generator A USAR 1

100 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 30 of49 4.4 Main Steam Line Break The Main Steam Line Break (MSLB) event is a rupture in the main steam piping resulting in a rapid depressurization of the SG secondary and corresponding cooldown of the primary.

The temperature reduction results in an insertion of positive reactivity with the potential for core power increase and DNBR violation.

The MSLB transient scenario presented here is modeled as an instantaneous, double-ended break at the nozzle of one steam generator from hot shutdown conditions with offsite power available. The input summary is provided in Table 4.4-1. Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.4-1 MSLB Input Summary I

I Parameter I Value I Notes I

Vessel T A V G (F) 547 I HZP nominal Initial Conditions Core power (MW)

Pump power (MW)

RCS Flow (gpmAoop)

I RCS Pressure (psia) 2250 I Nominal I

Pressurizer Level (%)

I 2 1 I HZP nominal I

1772.9E-9 8

89,000 I

SGLevel(%)

I 44 I Nominal I

HZP; USAR - 1.0% RTP Nominal Thermal Design Assumptions/Configuration Heat transfer option Manual Reactor Trip Main feedwater flow (% HFP value)

Auxiliary feedwater flow rate (gpm)

SG tube plugging (%)

Reactivity Parameters Boron Reactivity (pcmtppm)

I Decay heat multiplier I 1.0 I USAR= l.E-20 1 - Dominion method maximizes heat transfer coefficients for the faulted SG secondary side.

Doppler Reactivity Feedback Moderator Feedback Forced HT Map (note 1) 100 600 0

-8.0 USAR = Forced + Free Convection HT Map Assumed at time 0 sec initiated at time 0 sec initiated at time 0 sec; per SG Minimum value USAR= reactivity f(boron concentration).

Doppler Power defect, DTC model disabled Moderator density feedback USAR - Doppler power defect plus DTC Moderator density feedback

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 31 of49 Results - MSLB with Offsite Power Available The forcing function for the MSLB transient is the break flow. The main steam flow from the faulted and unfaulted SGs is plotted on Figure 4.4-1. Flow from the unfaulted SG stops at approximately 11 seconds due to Main Steam Isolation Valve (MSIV) closure. Flow from the faulted SG continues and there is good agreement between the Dominion and USAR cases for the first 85 seconds. After that time, the break flow predictions from the faulted steam generator begin to diverge slightly. This is primarily due to the differences in the predicted core heat flux response as discussed below.

The SG pressure response (Figure 4.4-2) matches well initially with the USAR data until the MSIV closes. After the MSIV closure, the unfaulted SG pressure increases and the USAR SG pressure remains higher than the Dominion model, most likely due to difference in SG modeling.

The core heat flux response is shown on Figure 4.4-3. The Dominion case core heat flux increases to a higher value compared to the USAR case. This results primarily from differences in the amount and timing of boron reaching the core. The Dominion boron transport method conservatively models the various system delays associating with purging of fluid from the SI lines and transport of borated water from the Refueling Water Storage Tank (RWST) to the core. The initial fluid in the SI piping is assumed to be at a boron concentration of zero ppm. This affects the timing at which boron reaches the core from the RWST and begins to suppress power. In addition, the USAR analysis RCS pressure decreases more quickly than the Dominion pressure. While this difference is relatively small, it allows for greater injection of accumulator flow for the USAR cases, which is further reflected in the power response as well as the RCS temperature and pressure response. Note that since accumulators are included in the KPS RETRAN model, the Dominion boron transport model described in Reference 1 was modified slightly to include the accumulators, however the accumulators are not subject to the system delays associated with the purge time for the Safety Injection (SI) system piping.

The total core reactivity is initially similar to the USAR data as shown in Figure 4.4-4.

After 60 seconds, the USAR core reactivity becomes negative, reflecting the sudden increase in core boron concentration. After approximately 100 seconds the negative reactivity for the Dominion model increases as the core boron concentration increases.

However, this occurs after the point of peak heat flux.

The core average boron concentration response is shown in Figure 4.4-5. As can be seen, boron starts to reach the core prior to 60 seconds for both the USAR case and the Dominion

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 32 of 49 case. In the USAR case, a higher accumulator flow rate results in a larger increase in core boron concentration. As seen on Figure 4.4-3 for the USAR analysis, the effect on core heat flux is immediate. Once accumulator flow ceases, core boron concentration continues to increase slowly as a result of continued SI flow from the RWST. In the Dominion model, the accumulator flow is less than the USAR case during the first 100 seconds of the transient. Due to the SI piping purge delay times in the Dominion model, borated water from the RWST does not enter the core until approximately 140 seconds. The higher core heat flux predicted by the Dominion case is mainly attributable to the later timing of boron injection.

The pressurizer pressure response is shown in Figure 4.4-6. The pressure initially decreases at a rate comparable to the USAR result. At approximately 20 seconds, the upper head begins to flash and the depressurization rate is decreased. The timing of the upper head flashing and the following depressurization is a contributor to when the accumulators activate. After the accumulator flow stops, the RCS pressure starts to rise slowly.

The reactor vessel inlet temperature response (Figure 4.4-7) shows that the initial cooldown matches well for both the USAR and Dominion cases. After approximately 150 seconds, the temperature differences are attributed to the different core heat flux response. However, this occurs well after the point of peak heat flux, as core power is steadily decreasing.

The sequence of events is compared to the USAR in Table 4.4-2. USAR values are taken from Kewaunee USAR Table 14.2.5-1.

Table 4.4-2 MSLB with Offsite Power Results Parameter Initiate Break Unfaulted SG High-High Steam flow setpoint reached Faulted SG Low-Low Steam Pressure Signal Unfaulted SG Low-Low Steam Pressure Signal Safety Injection Actuation Signal Steam line Isolation (MSIV Closure) occurs Peak Heat Flux occurs Feedwater Isolation occurs Peak Heat Flux (fraction of nominal)

Seauence of Events:

DOM 0.0 0.79 1.02 1.64 2.79 10.29 100.0 86.72 0.318 USAR 0.01 0.71 1.44 2.01 2.72 10.22 56.5 87.82 0.288

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 33 of49 Figure 4.4-1 MSLB w/ Offsite Power - Steam Flow 4500 1 Faulted - DOM

-Unfaulted

- DOM

-Faulted

- USAR

..-... Unfaulted - USAR a

Unfaulted SG Isolation 120 180 Time (seconds)

Figure 4.4-2 MSLB wl Offsite Power - Steam Generator Pressure 1 k

, Unfaulted SG lsolafion Faulted - DOM Unfaulted - DOM

-Faulted

- USAR Unfaulted - USAR Feedwater lsolatm 120 180 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.4-3 MSLB wl Offsite Power - Core Heat Flux pg. 34 of 49 120 180 Time (seconds)

Figure 4.4-4 MSLB w/ Offsite Power - Core Reactivity Causes USAR Reactlnty to Become Negat~ve 120 180 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.4-5 MSLB wl Offsite Power - Core Average Boron Concentration pg. 35 of 49 120 180 Time (seconds)

Figure 4.4-6 MSLB wl Offsite Power - Pressurizer Pressure USAR L...

120 180 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 36of 49 Figure 4.4-7 MSLB wloffsite Power - Reactor Vessel Inlet Temperature 120 160 Time (seconds)

Summary - MSLB This section presents a comparison of a RETRAN-02 Main Steam Line Break transient calculation with the Kewaunee model using the Dominion RETRAN transient analysis methods (Reference 1) compared to the USAR results. The key observations from these comparisons are that:

1) The peak power and heat flux reached with the Dominion methods is higher than the USAR result.

I

2) The effect of boron is significant in these transients because it can determine the timing and the magnitude of the transient peak power. The core average boron concentration resulting from SI and accumulator flow is different between the USAR and the Dominion model. The amount of flow from the accumulators greatly affects the amount of boron in the system and core power.

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 37 of49 4.5 Control Rod Bank Withdrawal at Power The Control Rod Bank Withdrawal at Power (RWAP) event is defined as the inadvertent addition of core reactivity caused by the withdrawal of rod control cluster assembly (RCCA) banks when the core is above no load conditions. The RCCA bank withdrawal results in positive reactivity insertion, a subsequent increase in core nuclear power, and a corresponding rise in the core heat flux. The RWAP event described here is terminated by the Reactor Protection System on a high neutron flux trip or the overtemperature AT trip (OTAT), consistent with the USAR analyses.

The RWAP event is simulated by modeling a constant rate of reactivity insertion starting at time zero and continuing until a reactor trip occurs. The Dominion analysis involves two different reactivity insertion rates, 3 pcrnlsec and 100 pcrnlsec that match the reactivity insertion rates described in the USAR. Both cases assume that the reactor is initially operating at 100% power, with minimum core reactivity feedback. Most of the input parameters are the same as those used in the USAR Chapter 14 analyses. Where differences from the USAR inputs exist, they are indicated in the Notes column.

I NSSS Power (MW) 1780 I Nominal plus pump heat Table 4.5-1 RWAP Input Summary Notes Parameter Initial Conditions RCS Flow (gpmlloop)

Vessel TAVG (F)

RCS Pressure (psia)

Value Pressurizer Level (%)

SG Level (%)

I Pressurizer level control I

I Not credited I

93,000 573 2250 Assumptions/Configur ation Reactor trip Automatic rod control Minimum Measured Flow Nominal Nominal 48 44 High neutron flux or OTAT Not credited Nominal Nominal Pressurizer heaters Pressurizer sprays, PORVs SG tube plugging (%)

Reactivity Parameters Doppler Temp. Coefficient (PC@)

Moderator Temp. Coefficient (PC@)

Moderator Density Coefficient Fuel Heat Conduction Model Initial Fuel Average Tem~erature 10

-1.2 0.0 N/A Minimum Not credited Active Max value Dominion least negative DTC model.

USAR uses least negative Doppler-only power coefficient, and a least negative DTC (driven by moderator temperature).

USAR uses a value of 0.0 pcd°F for MTC USAR uses 0.0 AW(gdcc)

USAR targets a minimum value

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 38 of49 Results - RWAP 3 pcdsec Case Figure 4.5-1 shows the core power response, which is slowly changing until a reactor trip occurs. The core power rate of increase for the Dominion case is somewhat greater than the USAR data. The Dominion case trips on high neutron flux at about 41 seconds, while the USAR case trips on an OTAT signal at about 45 seconds. It is noted that the USAR core power response very nearly reaches the 118% setpoint for the high flux trip. The difference in reactor trip mechanisms between the Dominion and USAR cases is reasonable, considering the breakpoint for switching between OTAT and high flux occurs at approximately 4 pcdsec, as shown in USAR Figure 14.1.2-3. The pressure response also affects the OTAT setpoint (setpoint will be lower in the USAR case, due to the USAR pressure response discussed below).

The pressurizer pressure response is shown in Figure 4.5-2. For the Dominion case, the pressure rises faster than the USAR result. For the first 5 seconds the results are similar.

However, the USAR result shows a flat pressure response from about 5 to almost 40 seconds maintaining the pressure at about 2255 psia. In the Dominion case, the pressure steadily increases rising above the USAR value and continuing until the reactor trips.

The RCS pressure response is determined by the modeling assumptions, especially pressurizer spray flow. As noted previously in Section 4.3, the Dominion method uses a conservative value for pressurizer spray flow rate; however, it appears that the USAR model adds additional conservatism, which suppresses the pressure increase associated with the RWAP event.

Figure 4.5-3 shows the RCS Loop A average temperature. There is good agreement between the USAR analysis and the Dominion model, as the peak temperatures are approximately 585 O F for the USAR, and 584 O F for the Dominion model. The time of peak temperature is related to the time of reactor trip as shown in Figure 4.5-1. The sequence of events for the 3 pcrnlsec RWAP transient is compared to the USAR in Table 4.5-2.

Table 4.5-2 RWAP 3 pcdsec Time Sequence of Events I

I Reactivitv Insertion at 3 ocdsec 0.0 I

0.0 I

Event I Reactor Trio Signal Initiated 41.27" 45.28""

Time (seconds)

• Trip on high neutron flux

• • Trip on OTAT now USAR

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 39 of 49 Results - RWAP 100 pcmlsec Case Figure 4.5-4 shows the core power response which rises rapidly until a reactor trip on high flux occurs. The Dominion case trips on a high neutron flux signal of 118% at about 1.8 seconds, compared to about 2.03 seconds for the USAR case (each includes a 0.65 second delay). The Dominion case includes decay heat while the USAR analysis neglects decay heat. The effect of decay heat modeling differences is seen post-trip, where the USAR case power drops to the delayed neutron stable period following shutdown, while the Dominion case follows a decay heat curve defined by the ANS-5.1 1979 single isotope (U-235) model.

The 100 pcm/sec transient is a fast transient and the time period before the reactor trip is so brief that the any differences in fuel pin heat transfer modeling assumptions have little impact on Doppler reactivity feedback.

The pressurizer pressure response is shown in Figure 4.5-5. As is the case with the RWAP analysis for a 3 pcm/sec reactivity insertion rate, the Dominion model predicts a higher pressurizer pressure response. This is mainly due to the differences in modeling of the pressurizer spray.

Figure 4.5-6 shows the RCS Loop A average temperature. There is good agreement between the USAR analysis and the Dominion model, as the peak temperatures are approximately 576 OF for both models. The sequence of events for the 100 pcmlsec RWAP transient is compared to the USAR in Table 4.5-3.

Table 4.5-3 RWAP 100 pcrnlsec Time Sequence of Events Event Time (seconds)

DOM USAR Reactivity Insertion at 3 pcmlsec Reactor Trip Signal Initiated

• Trip on high neutron flux 0.0 1.78" 0.0 2.03"

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.5-1 RWAP 3 pcmlsec - Core Power pg. 40of49 1

-DOM USAR 40 50 60 70 80 90 100 Time (seconds)

Figure 4.5-2 RWAP 3 pcmlsec - Pressurizer Pressure 0

10 20 30 40 50 60 70 80 90 100 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 41 of 49 Figure 4.5-3 RWAP 3 pcmlsec - RCS Average Temperature 40 50 60 70 Time (seconds)

Figure 4.5-4 RWAP 100 pcmlsec - Core Power 3

4 5

6 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.5-5 RWAP 100 pcmlsec - Pressurizer Pressure pg. 42 of49 DOM USAR 3

4 Time (seconds)

Figure 4.5-6 RWAP 100 pcmlsec - RCS Average Temperature 3

4 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 43 of 49 Summary - RWAP The Dominion Kewaunee analysis provides results that are similar to the USAR analysis for the RWAP event. There are small differences in core power response during the early portion of the transient. For lower reactivity insertion rates, the fuel pin heat transfer modeling differences can affect the time to reactor trip; however, the peak core powers and peak average coolant temperatures are in close agreement. The USAR spray flow rate appears to include additional conservatism, resulting in a less severe pressurizer pressure response compared to the Dominion analysis.

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 44 of49 4.6 Loss of Flow The Loss of Flow (LOF) event is the loss of one or two reactor coolant pumps (RCP) and an associated coastdown of reactor flow. The reduction in core flow results in a sudden increase in coolant temperature and increased probability of violating a DNBR limit.

The LOF transient scenario presented here is a complete loss of flow resulting from the trip of both RCPs. The input summary is provided in Table 4.6-1. Where differences from USAR inputs exist, they are indicated in the Notes column.

Table 4.6-1 LOF Input Summary I

I Parameter 1 Value 1 Notes I

Pressurizer Level (%)

I 48 I Nominal I

Initial Conditions NSSS Power (MW)

RCS Flow (gpdloop)

Vessel T A V C (F)

RCS Pressure (psia) 1780 93,000 573 2250

, r SG Level (96)

SG Pressure (psia)

Nominal plus pump heat Minimum measured flow Nominal Nominal Automatic rod control Doppler Temp. Coef (pcm/F)

Assumprlon!

Pum- '- --'

Rea~,"' uLp I

1 u u v v L \\ b u I*uw li) w b u v e I

44 804.5 I Not credited Main steam dumps, SG PORV SG tube plugging (%)

Reactivity Parameters USAR uses most negative Doppler-only power coefficient, and a least negative DTC (driven Nominal Pressurizer swavs. PORVs 10 Results - LOF I Not credited Not credited Max value Moderator Temp. Coef RCS flow coasts down following LOF as shown on Figure 4.6-1. The Dominion results compare well with the USAR data demonstrating good agreement between the pump model and friction losses. Since there is minimal reactivity feedback, the core power remains nearly constant prior to the reactor trip on low RCS flow, and the Dominion and USAR power response compares well as seen on Figure 4.6-2. The resulting core average heat flux is provided on Figure 4.6-3. The Dominion case provides a higher, more conservative value 0

by moderator temp).

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 45 of 49 than the USAR case. This is primarily due to the fact that the Dominion model is initialized at a higher fuel temperature with higher stored energy. The USAR assumes a higher gap conductance and U02 thermal conductivity resulting in lower initial fuel temperature and stored energy. A sensitivity case is also shown on Figure 4.6-3 that modifies the inputs to more closely match the US AR case initial conditions and assumptions, including the higher gap heat transfer conductance and U02 thermal conductivity multiplier. In this case, the heat flux closely tracks the USAR value.

The RCS loop temperature response is shown on Figure 4.6-4. Again, the higher stored energy results in higher loop temperatures for the Dominion case compared to the USAR response. This is also reflected in the pressurizer pressure response shown on Figure 4.6-5 where the RCS pressure for the Dominion case peaks above the USAR case and remains higher for the duration of the event. The effect of pressurizer liquid flashing after the pressure peak can be seen by a reduction in the rate of pressure decrease.

Table 4.6-2 LOF Results Summary - LOF Parameter Reactor Trip (sec)

(Low RCS flow)

The Dominion Kewaunee LOF analysis provides RCS flow response that is very similar to the USAR results, demonstrating close agreement between the pump model and friction losses. The timing for reactor trip and the core power response is also in close agreement.

The core heat flux is higher for the Dominion case due to differences in initial fuel temperature and stored energy. This is also reflected in higher primary side pressure and temperature values. The higher heat flux is conservative for DNBR acceptance criteria.

Seauence of Events:

DOM 2.6 1 USAR 2.57

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 46 of 49 Figure 4.6-1 LOF -Total Core Inlet Flow 460 1

4 6

8 Time (seconds)

Figure 4.6-2 LOF - Nuclear Power 4

6 8

Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information Figure 4.6-3 LOF - Core Average Heat Flux pg. 47 of49 4

6 8

Time (seconds)

Figure 4.6-4 LOF - RCS Loop Temperature 520 1 0

2 4

6 8

10 12 Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 48 of49 Figure 4.6-5 LOF - Pressurizer Pressure 2360 USAR I 4

6 8

Time (seconds)

DOM-NAF-5 Attachment B RETRAN Benchmarking Information pg. 49 of 49 5.0 Conclusions This report presents demonstration transient analyses performed with the KPS RETRAN model developed in accordance with VEP-FRD-41. These analysis results are compared with current Kewaunee USAR results. The following conclusions are drawn based on these analyses.

1) This report demonstrates that the Dominion RETRAN-02 model and analysis methods can predict the response of transient events with results that compare well to US AR results.

2) Where there are differences between the Dominion results and the USAR results, they are understood based on differences in noding, inputs, or other modeling assumptions.

3) The Dominion Kewaunee RETRAN-02 model is consistent with current Dominion methods (Reference I). These methods have been applied extensively for Surry and North Anna licensing, engineering and plant support analyses.

4) The RETRAN comparison analyses satisfy the DOM-NAF-5 applicability assessment criteria and provide further validation of the conclusion that Dominion's RETRAN analysis methods are applicable to Kewaunee and can be applied to Kewaunee licensing analysis for reload core design and safety analysis.

6.0 References

1)

Topical Report, VEP-FRD-41, Rev. 0.1-A, "VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code," June 2004.

2)

Topical Report, VEP-NFE-2-A, "VEPCO Evaluation of the Control Rod Ejection Transient," December 1984.