ML15261A070

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Forwards Nonproprietary & Proprietary Handouts Presented in 911007 & 08 Meeting W/Nrc & Contract Reviewers to Discuss Outstanding Issues Associated W/Topical Repts DPC-NE-2004, DPC-NE-3000 & DPC-NE-3001.Proprietary Handouts Withheld
ML15261A070
Person / Time
Site: Oconee, Mcguire, Catawba, McGuire  Duke Energy icon.png
Issue date: 10/16/1991
From: Tuckman M
DUKE POWER CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML15261A071 List:
References
NUDOCS 9111060338
Download: ML15261A070 (188)
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Text

ATTCHEI g ATTACHMENT 0e

NRC/DUKE MIEETING M1C8 RELOAD & DUKE TOPICAL REPORTS October 7&8, 1991 I. RESPONSES TO SPECIFIC NRC OUESTIONS

1. BWCMV Applicability to OFA
2. Validity of MAP limits, AD limits, FAH limits to OFA
3. Barton Transmitter Error
4. Reactor Vessel Flow Anomaly
5. VIPRE SCD Methods and Applications to Transients
6. UCBW @ HZP Peaking and Statepoints Used in SCD
7. SG Narrow Range Level Modeling
8. Pressurizer Inter-Region Heat Transfer
9. Control System Modeling Philosophy II. McGUIRE/CATAWBA RETRAN MODEL NODALIZATION AND JUSTIFICATION
1. Overview
2. SG Secondary
3. Reactor Vessel Head and Core
4. Transient-Specific Deviations from Generic Modeling III. DETAILED DISCUSSION OF TRANSIENT MODELING AND RESULTS WITH COMPARISON TO CURRENT FSAR ANALYSES
1. Increase in Heat Removal - No presentation planned since NRC has already reviewed steam line break in DPC-NE-3001
2. Decrease in Heat Removal - Turbine Trip
3. Decrease in Flow - Locked Rotor
4. Reactivity Transient - Uncontrolled Bank Withdrawal at Power
5. Loss of Inventory - No presentation planned since NRC has already reviewed the Catawba SGTR analysis
6. Summary IV. DPC-NE-3002 NRC OUESTIONS (PRELIMINARY DISCUSSION)

V. ADDITIONAL DISCUSSION VI. IDENTIFICATION OF REMAINING ISSUES ON MIC8 RELOAD VII. OCONEE RETRAN MODEL NODALIZATION AND JUSTIFICATION

1. Overview
2. Reactor Vessel Head and Core
3. Steam Generator
4. RETRAN Model Validation Summary
5. Comparison to Other Nodalizations
6. Nodalization Sensitivity Studies

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DPC-NE-3000 TOPICAL REPORT CHRONOLOGY OF FORMAL TRANSMITTALS

1. SEPTEMBER 29, 1987: DUKE LETTER SUBMITTING DPC-NE-3000.
2. APRIL 3, 1989: DUKE LETTER DOCUMENTING PHONE DISCUSSIONS WITH NRC/ITS. PROPOSED REVISING ONS VIPRE CHAPTER, DELETING CHAPTER 6, AND ADDING A McGUIRE/CATAWBA VIPRE CHAPTER.
3. APRIL 7, 1989: NRC LETTER AGREEING TO REVISION 1 SCOPE CHANGES (ITEM 2), AND FORWARDING SET OF QUESTIONS RELATED TO OCONEE RETRAN.
4. MAY 11, 1989: DUKE LETTER SUBMITTING REVISION 1.
5. JUNE 13, 1989: NRC LETTER FORWARDING SET OF QUESTIONS RELATED TO McGUIRE/CATAWBA RETRAN.
6. JUNE 15, 1989: DUKE LETTER WITH RESPONSES TO NRC QUESTIONS ON OCONEE RETRAN (ITEM 3).
7. JUNE 19, 1989: DUKE LETTER DESCRIBING DIFFERENCES BETWEEN THE DPC-NE-2003 AND DPC-NE-3000 VIPRE MODELS.
8. JULY 25, 1989: NRC LETTER FORWARDING SET OF QUESTIONS RELATED TO VIPRE MODELS.
9. AUGUST 9, 1989: DUKE LETTER DESCRIBING OTSG MODELING.
10. SEPTEMBER 13, 1989: DUKE LETTER RESPONDING TO NRC QUESTIONS ON McGUIRE/CATAWBA RETRAN (ITEM 5).
11. FEBRUARY 20, 1990: DUKE LETTER RESPONDING TO NRC QUESTIONS ON VIPRE MODELS (ITEM 8).
12. AUGUST 3, 1990: DUKE LETTER RESPONDING TO TWO UNDOCUMENTED

- NRC QUESTIONS RELATED TO VIPRE MODELS.

13. AUGUST 29, 1991: DUKE LETTER RESPONDING TO PROPOSED NRC SER RESTRICTIONS RELATED TO DPC-NE-3000 AND DPC-NE-2004.
14. SEPTEMBER 25, 1991: DUKE LETTER SUBMITTED A NEW PROPRIETARY AFFIDAVIT PER NRC REQUEST.
1. Jan. 9, 1989, Duke Letter to NRC, submittal of DPC-NE-2004
2. Feb. 22, 1990, Duke letter to NRC, revisions to topical report
3. Aug. 2, 1990, NRC letter to Duke, 1st round of questions
4. Sept. 14, 1990, Duke letter to NRC, response to 1st round of questions
5. Nov. 29, 1990, Duke letter to NRC, response to questions from conference call
6. Aug. 29, 1991, Duke letter to NRC, response to final set of questions
7. Sept. 25, 1991, Duke letter to NRC, update affidavit

Use of BWCMV CHF Correlation for OFA Fuel

  • The BWCMV correlation is based on an extensive data base with parameters that span the OFA parameters, although specific OFA data is not included.
  • Two important fuel design parameters are hydraulic diameter and rod-to-rod gap
  • The OFA hydraulic diameter (0.5 10 in.) and rod-to-rod gap (0.136 in.) are within the range of the BWCMV data base.

I'

  • BWCMV results given in BAW-10159P-A show negligible trend or bias to hydraulic diameter and rod-to-rod gap.
  • This led the NRC to conclude in the SER that the BWCMV correlation applies to OFA fuel.

VIPRE-01/BWCMV

  • VIPRE-O1/BWCMV and LYNX2/BWCMV results have been compared to show that both codes give essentially identical results.

N Mean M/P Std. Dev.

VIPRE-01 124 0.9962 0.1045 LYNX2 124 1.0006 0.1082

Measured/Predicted CHF Vs.

Hydraulic Diameter 1.4 t

T 1.2

-o X LYNX2 oVIPRE-01 C,,

Soa

~0.8 0.6 0.35 0.4 0.45 0.5 0.55 Hydraulic Diameter - Inches OO

VIPRE-01 vs. LYNX2 DNBR BWCMV CHF CORRELATION 1.4 z

0.1 0.6 0.6 0.7 0.8 0.0 1 1.1 1.2 1.3 1.4 1.5 1.6 VIPRE-01 DNBR

0 Mixed Core Modelling

. Mixed core analyses specifically model both the OFA and MKBW fuel. The peaking associated with each fuel type is explicitly calculated and compared against MAP limits.

. . MAP limits are the total allowable peak that would reproduce the DNBR limit functionalized against the magnitude and location of the axial peak.

  • . Mixed core MAP limits apply to both OFA and MKBW fuel.

. AFD limits are set based on the allowable peaking as dictated by the MAP limits.

  • Mk-BW MAP limits are applicable to OFA fuel in a transition core
  • MAP limits are calculated for a full Mk-BW core based on a DDL of 1.55
  • A conservative transition core DNBR penalty for OFA fuel of 3.8 % is applied against the magin included in the DDL
  • The SDL for OFA fuel is virtually identical to the SDL for Mk-BW fuel N SDL Mk-BW 3000 1.337 OFA 3000 1.322

McGuire DNBR Penalties

  • Statistical DNBR Limit (SDL) = 1.40
  • Design DNBR Limit (DDL) = 1.55
  • DNBR Margin = 10.7 %

Transition Core DNBR Penalty

  • Optimized Fuel Assembly (OFA) total AP is 2.4 %>

Mark-BW fuel assembly AP

  • Flow will be forced out of the OFA fuel into the Mk-BW fuel
  • A generic transition core analysis was performed to determine a DNBR penalty for the OFA fuel

PROPRIETARY INFORMATION NOTICE THE ATTACHED DOCUMENT CONTAINS OR IS CLAIMED TO CONTAIN PROPRIETARY INFORMATION AND SHOULD BE HANDLED AS NRC SENSITIVE UNCLASSIFIED INFORMATION.

IT SHOULD NOT BE DISCUSSED OR MADE AVAILABLE TO ANY PERSON NOT REQUIRING SUCH INFORMATION IN THE CONDUCT OF OFFICIAL BUSINESS AND SHOULD BE STORED, TRANSFERRED, AND DISPOSED OF BY EACH RECIPIENT IN A MANNER WHICH WILL ASSURE THAT ITS CONTENTS ARE NOT MADE AVAILABLE TO UNAUTHORIZED PERSONS.

COPY NO.

DOCKET NO. _

CONTROL NO.

REPORT NO.

REC'D W/LTR DTD.

NRC Form 190 (10-89)

NRCM 2101 PROPRIETARY INFORMATION

PROPRIETARY MARK-BW vs. OFA PRESSURE DROP 26 24 Support Plate Upper Nozzle 221 End Grid OFA 20d Mark-BW (w/Debris-Filtering Bottom Nozzle)

~16~

Ca. 1 g 14A CL

, 121 a) 101 o0 8 -~Mixing Grid M Mixing Grid 6

Non-Mixing Grid 41 End Grid Bottorn Nozzle 2

Support Plate 0

20 40 60 80 100 120 140 160 180 Axial Location, in 9111060347 911016 PDR ADOCK 000026.0 CF L

  • For the transition cycles containing Mk-BW & OFA fuel, the core safety & operating limits are based on analyses of a full Mk-BW core as described in DPC-NE-2004
  • A conservative transition core VIPRE-01 model was developed consisting of one OFA assembly (the hot assembly) and the rest of the core is modeled as Mk-BW assemblies

75 CHANNEL MARK-BW/OFA VIPRE-01 MODEL 1-45 OFA k-BWW Dk-

  • k-BW 48 49D k-B14 KBW Mk-B1 Mk-BW 51 Uk-BW 2

Mk-BW Mk-BW SMk-BW 53 Mk-BWHko.

Mk-BW 54 Di-B x do1 FA Design zk-BW k-BW Mk-BW Ilk-BW 60 st6 2 63 Mk64

-BW IMk-65\ChneNo BW 7-BW 66 7 MkSk-BW 68 W Mk-BW 69 Mk-BW 70 Mk-BW Mk - BW IMk-BWI M1k-8W Mki -BW 72 -3 74 75

The following conditions were analyzed:

  • core safety limit statepoints
  • inlet & outlet skewed axial power shapes &

corresponding max. allowable radial peaks

0 0 The transition core MDNBR is compared with the corresponding MDNBR for a full Mk-BW core to determine the transition core penalty Transition Core DNBR Penalty = Mk-BW DNBR - Transition Core DNBR Mk-BW DNBR

0 0I

  • OFA Transition Core DNBR Penalty = 3.8 %
  • The generic transition core penalty is applied against the margin (10.7 %) included in the Design DNBR Limit (DDL) of 1.55

Instrumentation DNBR Penalty

  • The instrumentation penalty accounts for measurement biases that are constant in direction and magnitude.
  • McGuire instrumentation biases:
  • Pressure (Barton transmitters)
  • RCS flow (Barton trans. & FW venturi fouling)
  • The instrumentation penalty is based on individual parameter sensitivity studies.
  • Total instrumentation DNBR penalty = 1.9 %

(McGuire)

Flow Anomaly

  • Several Westinghouse 4-loop plants have observed an anomalous condition concluded to be an aperiodically occurring vortex flow disturbance in the lower plenum.
  • Data was taken at 19 plants to determine if the anomaly was present or not (WCAP-1 1528, April 1988)
  • The flow anomaly does occur at both Catawba units, but is not observed at either of the McGuire units.
  • A DNBR penalty was calculated to account for the flow anomaly at Catawba.

Catawba Flow Anomaly

  • The flow anomaly core inlet flow distribution was determined based on Catawba core exit thermocouple and power distribution data.
  • The DNBR penalty was calculated in the same manner as the transition core penalty, comparing the flow anomaly inlet flow distribution DNBR results with the generic DNBR results

McGuire DNBR Penalties DNBR Penalty Mk-BW OFA Transition Core 0 3.8 %

Instrumentation/Hardware 4.7 % 1.9 %

Rod Bow 0 3.5 %

Flow Anomaly 0 0 Total DNBR Penalty 4 .7  % 9.2 %

Available DNBR Margin 6.0 % 1.5 %

Transient Barton Pressure Transmitter Bias Modeling

  • Transients which are analyzed for peak RCS pressure are not affected since a high initial pressure is conservative and the bias is a pressure reduction
  • Transients which are analyzed for peak Main Steam System pressure are bounded by turbine trip. This transient is not affected since a high initial pressure is conservative and the bias is a pressure reduction
  • DNB transients which use the SCD methodology account for the bias in the SDL e DNB transients which use a non-SCD methodology adjust the pressurizer pressure initial condition down by an additional amount to bound the effect

VIPRE-01 SCD Methods and Application to Transients

  • The Statistical Design Limit (SDL) of 1.40 calculated in DPC-NE-2004 is conservative for all conditions where the SCD methodology is used.

VIPRE-01 SCD Methods

  • The "Simplified" or Brute Force method runs the VIPRE-01 code instead of using the Response Surface Model (RSM) to calculate DNBR.
  • FIGURE A depicts the analysis flow path using the RSM. HGURE B shows the same process with the Simplified method.

011.

FIGURE A RSM BASED S C D Develop Matrix of (

VIRE 78 Cases)

S Run VIPRE~ Cases Deve op RSM

. .(..... 36 Coefs.

Determine Uncertainties Core Power, Tin, RC Flow, ByasFoP ee ob... 4 ....

jFA\ H-measurement, FESH, FtA H Spacing, FZ, Z, RSM, Code/Model, CHF Correlation 10 Propagate eper. Uncertainties Statepoints (Monte Carlo - RSM J Determine SCD Limit and Design DNBR Limit (DDL)

To Applications

FIGURE B "SIMPLIFIED" S C D Core Power, Tin, RC Flow, Determine Uncertainties Bypass Flow, Pressure, FL H-measurement, F AH, FL HSpacing, FZ, Z, Code/Model, CHF Correlation Oper. Propagate Uncertainties Statepoints (Monte Carlo - VIPRE)

Determine SCD Limit and Design DNBR Limit (DDL)

IA To Applications

TABLE 1 The following values are from the simplified method analysis of the statepoint for the FSAR Section 15.3.2 transient, Complete Loss of RCS Flow.

VIPRE-01 Case Power RCS Flow Tcold Pressure FdelH Fz Z MDNBR 0 95.7% 71.79% 561.9 F 2320.0 # 1.500 1.550 0.70 1.601 2 95.9% 70.18% 564.8 F 2325.5 # 1.515 1.528 0.67 1.519 10 96.9% 71.38% 561.8 F 2328.2 # 1.489 1.473 0.69 1.731 100 95.6% 71.40% 564.0 F 2330.0 # 1.460 1.561 0.66 1.787 250 94.5% 74.65% 563.0 F 2329.0 # 1.539 1.575 0.72 1.467 500 95.2% 71.67% 563.9 F 2301.8 # 1.463 1.594 0.68 2.000 The SDL is calculated as described in DPC-NE-2004.

SDL= 1 / [1- (Chi Square

  • Kfactor
  • CV)]

where Mean = 1.619 Kfactor = 1.763 Stand. Dev = 0.2119 - Chi Square = 1.055

  • Coeff. of Variation = 0.1309 SDL = 1.322 There are no differences in the method between the 500 and 3000 case runs.

Application To Transients

  • The application of the SCD methodology to a transient is determined by the conditions at the point of MDNBR.

Table A. SCD Transient Limiting Statepoints Core Core Inlet Core Power Inlet Flow Temperature Pressure Basis FSAR (% FP) (%)* (OF) (psia) _F4,_ , Z for SDL (RSM range)(70-125) (80-125) (529.5-609.5) (1800-2455) 15.2.8 95.7 77.4 571.7 2302 1.50 1.55 0.7 Simplify 15.3.1 98.6 89.3 561.7 2301 1.50 1.55 0.7 RSM 15.3.2 95.7 71.2 561.9 2320 1.50 1.55 0.7 Simplify 15.4.1 80.8 69.8 558.4 2444 1.725 1.55 0.7 Simplify 15.4.2

a. 100 % 111.6** 97.8 566.4 2250 1.50 1.55 0.7 RSM 112.7 97.5 566.6 2318 1.50 1.55 0.7 RSM
b. 50 % 107.3 96.7 568.0 2317 1.725 1.55 0.7 RSM
c. 10 % 105.9** 95.5 572.9 2254 1.725 1.55 0.7 RSM 94.5 95.8 570.8 2370 1.725 1.55 0.7 RSM 15.4.3
a. Dropped 111.7 99.0 560.0 2311 1.574- 1.316- 0.7 RSM Rod 1.700 1.335
b. Single Rod 112.5 96.8 571.1 2318 1.50 1.55 0.7 RSM Withdrawal
c. Statically 100.0 98.4 561.6 2280 1.50 1.55 0.7 RSM Misaligned Rod 15.6.3 bounded by 15.2.8
  • % of nominal core inlet flow Nominal core inlet flow = (1-0.06)(382,000 gpm)

= 359,080 gpm

    • two different reactivity insertion rates

SZ PROPRIETARY DUKE POWER CO.

Table B. Monte Carlo Uncertainty Propagation Results (VIPRE-01)

Standard Coefficient Case N Mean Deviation of Variation SDL*

15.2.8 DPC-NE-3000 VIPRE-01 model 500 1.623 0.2109 0.1300 1.319 15.3.2 DPC-NE-3000 VIPRE-01 model 500 1.619 0.2119 0.1309 1.322 15.4.1 DPC-NE-3000 VIPRE-01 model 500 1.859 0.2334 0.1256 1.305 DPC-NE-2004 VIPRE-01 model 500 1.914 0.2381 0.1244 1.301 DPC-NE-3000 VIPRE-01 model 3000 1.846 0.2409 0.1305 1.291 DPC-NE-2004 VIPRE-01 model 3000 1.901 0.2458 0.1293 1.288

  • SDL Limit = 1.40 (DPC-NE-2004)

Application To Transients

  • The Statistical Design Limit of 1.40 determined in DPC-NE-2004 conservatively bounds all SCD transients.

UCBW From Zero Power Peaking Analysis:

. A three dimensional peaking analysis is performed at the MDNBR statepoint.

. Power distributions are evaluated for the two highest worth (differential and integral) sequential control banks moving in 100% overlap. Fully inserted to fully withdrawn bank positions are analyzed.

. The resulting power distributions are compared to the allowable peaking provided by the MAP limits.

. Acceptable results are no DNB.

. Typical maximum radial and axial peaks are less than 2.25 and 3.0, respectively.

UCBW From Zero Power Peaking Analysis cont'd:

In the event that negative DNB margins are found, the following analyses would be performed.

a. Analyze the suspect shape(s) in VIPRE, or
b. Reanalyze the UCBW from zero power with cycle-specific values, or
c. As a last resort, redesign the core.

UCBW AT Zero Power SCD Statepoints The SDL for the UCBW transient from Zero Power is calculated for two peaking conditions. The first condition is the transient statepoint. The second condition is the largest radial and axial peaking permitted by the MAP limits.

Condition Power RCS Flow Tcold Pressure FdelH Fz Z 1 80.8% 69.80% 558.4 F 2444.0 # 1.725 1.550 0.70 2 80.8% 69.80% 558.4 F 2444.0 # 2.256 2.100 0.01 A 500 case simplified method analysis was performed for each condition.

The results are as follows:

Standard Coefficient Condition Mean Deviation of Variation SDL 1 1.914 0.2381 0.1244 1.301 2 1.596 0.1798 0.1127 1.267

O Steam Generator Narrow Range Level Modeling

  • Plant instruments are differential pressure transmitters measuring the AP between two pressure taps located 233" apart in the upper downcomer
  • Plant procedures calibrate the instruments such that 0.3204 psid is equivalent to 0%

indicated level and 6.2142 psid is equivalent to 100% indicated level

  • The indicated level output at McGuire has an electronic lag of 1.5 seconds. This lag might be installed on the Catawba instrument in the future.

0

.. ,,Stearn outlet To Turbine Secondary Upper Wide Range Moisture

'Level Tap Separator Upper Narrow Range Manway Level Tap Separator Lower Narrow Range Anti-Vibration Level Tap a ra Tube Sundle Tube Support Plate(s)

Pre-heat Area

-- ,eFeedwater Inlet Lower Wide Range Tube Sheet Level Tap SPLIT FLOW PREHEAT Coolant Manway STEAM GENERATOR Inlet McGUIRE NUCLEAR STATION FIGURE 5.5.2-1

PROPRIETARY DUKE POWER CO.

  • The pressure difference between two downcomer junctions, each located at the elevation of the corresponding plant pressure tap, is calculated
  • This pressure difference is corrected for the static head difference between the mixture level in a right circular cylinder and the mixture level in the actual down comer
  • The corrected pressure difference is used to interpolate on a general data table containing the plant calibration values to give an indicated level signal
  • The response of the level model is generally good, as shown by the following figure. Occasional spiking occurs and is assessed for nonconservatism on a case-by-case basis

MNS-2 PARTIAL LOSS OF MAIN FEEDWATER 6/24/85 EVENT 1e A. a S

6 N

R L

E 4 V

E L

28 e 1ee 200 300 400 588 688 708 888 gee TIHE(SECONDS)

DATA - RETRAN

  • C

PROPRIETARY L DUKE POWER CO.

Pressurizer Interregion Heat Transfer Coefficient

  • Transients which are analyzed for peak RCS pressure model no interregion heat transfer since such heat transfer would nonconservatively reduce pressure
  • Transients for which the pressurizer sprays or PORVs are actuated are insensitive to the interregion heat transfer assumption since these pressure control device effects are dominant
  • Transients of very short duration do not allow time for significant amounts of heat to be transferred
  • For transients which do not fit into one of these categories a sensitivity study is done, e.g., UCBW core cooling capability analyses with pressurizer pressure inoperable

Control System Modeling Philosophy

  • Control systems are assumed to function only if their operation will make the transient more severe or if such operation is necessary for the transient to occur at all. No credit is taken for control system
  • operation which makes the transient less severe.
  • Control systems are generally assumed to operate with nominal gains, setpoints, and time constants. Complicated control systems are sometimes modeled conservatively.
  • Process parameter indications which are inputs to control systems are conservatively adjusted to account for
  • instrument uncertainties.
  • Control System Modeling Philosophy Example
  • As an example of these control system modeling principles, consider the dropped rod transient:
  • Instrument uncertainty is assumed on the rod position indication
  • Increased attenuation of neutron flux due to downcomer cooldown is modeled
  • A single failure is assumed in the Rod Control System auctioneering logic, causing it to select the lowest power range neutron flux channel as the power input
  • Control System Modeling Philosophy Example
  • No credit is taken for the high neutron flux rod withdrawal block (set at 103 % RTP) in mitigating the transient
  • If the rod withdrawal block
  • were operable the transient would not be a DNB concern since the cessation of rod reactivity addition at 103 %

power allow the thermal feedback to limit heat flux to a value commensurate with the available flow, while the pressure and temperature values would be less limiting than at 100% power

McGUIRE/CATAWBA RETRAN MODEL NODALIZATION AND JUSTIFICATION OVERVIEW o EFFORT BEGAN IN 1981 o NON-LOCA APPLICATIONS o BASE MODEL NODALIZATION TO BE BROADLY APPLICABLE o INFLUENCED BY RETRAN COMMUNITY

- UNIMPRESSED WITH SIMPLIFIED SG SECONDARY MODELS IN USE

- LITTLE VALIDATION BEYOND FSAR COMPARISONS o CONSTRAINED BY RETRAN CODE CAPABILITIES AND OPTIONS

- ONE DIMENSIONAL

- HEM (WITH SPECIAL PURPOSE MODELS)

- BEST POSSIBLE MODEL DESIRED

- AVOID EXCESSIVE DETAIL/COST o FOLLOW EARLIER OCONEE RETRAN APPROACH

NODALIZATION APPROACH o NODAL BOUNDARIES BASED ON

- PHYSICAL BOUNDARIES

- FLOW PATTERNS

- PHASE SEPARATION

- THERMAL GRADIENTS

- CODE LIMITATIONS o SYMMETRICAL LOOPS WITH COMPLETE MIXING IN THE REACTOR VESSEL o USE OF THE TRANSPORT DELAY MODEL o RETRAN COMMUNITY NORMS o COMPUTING COST o EXPERIENCE GAINED WITH INITIALIZATION AND MATCHING PLANT DATA (BENCHMARKING)

Fizure 11-6 McGuire RETRAN Model Nodalization Diagram PR RIT Y DUKE POWER CO.

jig? 1169 J11 ji166 J17931~

JES J1?f

'113l jig5 JoelVI V15315 -lls VI V11IJ9 _ _ _ _

vis 105 V J151 ila Jimits it vi1a@visit14 i is-14aV

PROPRIETARY DUKE POWER CO.

STEAM GENERATOR NODALIZATION OBJECTIVES o GOOD REPRESENTATION OF THE PREHEATER o EXPLICIT MODELING OF RECIRCULATION o LEVEL INDICATIONS o MATCH LIQUID INVENTORY DATA o LIQUID DISTRIBUTION REASONABLE o HEAT TRANSFER NORMALIZATION REASONABLE o TRANSIENT PERFORMANCE MATCHES DATA CONSTRAINTS o RETRAN PHASE SEPARATION CAPABILITY o PERFORMANCE BASELINE UNCERTAINTY

Split Flow Preheater Steam Generator U E Model *D2/D3 Str .!e Setoncary Upper Wide Range hlosture Level Tap Separator aye Upper Narrow Range Level Tap______

Pri Dmry Separator F

_hi_

Lower Narrow Range Anti-Vibration Level Tap eats Tube EL; ~e 0111 Feedwater Ii 1***L**Inlet Lower Wide Range ___ii i Tube Sheet Level Tap 0

Cc lant Manway- -

Split Flow Preheat Steam Generator Mc~uire Unit I Mic~uire Unit 2 Catawba Unit I 2-11

PROPRIETARY RESULTS DUKE POWER CO.

o 8 PRIMARY VOLUMES o 7 SECONDARY VOLUMES o 8 SG TUBE CONDUCTORS o PASSIVE CONDUCTORS o PREHEATER o RECIRCULATION o LEVEL INDICATIONS o PHASE SEPARATION o INITIALIZATION PROCESS

- MASS AND LEVEL MATCHED

- HEAT TRANSFER NORMALIZED BY ADJUSTING PRESSURE o TRANSIENT BENCHMARKS VALIDATE MODELING

51 RETRAN MODEL VALIDATION OBJECTIVES o STEADY-STATE

- FULL POWER

- PARTIAL POWER

- NATURAL CIRCULATION o FLOW COASTDOWNS o SG HEAT TRANSFER o PRESSURIZER PHENOMENA o POINT KINETICS CORE MODEL o TRANSIENT SPECTRUM

- OVERCOOLING

- UNDERCOOLING

- LOSS OF FLOW

- REACTIVITY EVENTS

- LOSS OF INVENTORY

  • BENCHMARK TRANSIENTS o STEADY-STATE AT POWER o STEADY-STATE NATURAL CIRCULATION o RCP COASTDOWN TESTS (4 CONFIGURATIONS) o LOSS OF MFW o PARTIAL LOSS OF MFW o STEAM LINE PORV FAILURE o ASYMMETRIC NATURAL CIRCULATION o SINGLE RCP TRIP o LOSS OF OFFSITE POWER o TURBINE TRIP

PHENOMENA/BEN4ARK MATRIX Steady State 5.1.1 5.1.2 5.2.1 5.3.1 5.3.2 5.3.3 5.3.4 5.4.1 o SG HEAT TRANSFER X X X X X X X X o SG LEVEL INDICATION X X X X X X X o AFW HEAT TRANSFER X X X X X o EXCESSIVE STEAMING X X o PRESSURIZER INSURGE X X X X o PRESSURIZER OUTSURGE X X X X X X o PRESSURIZER EMPTYING X o PRESSURIZER SPRAY/PORVs X o POINT KINETICS/THERMAL X X FEEDBACK o FLOW COASTDOWN X X X o NATURAL CIRCULATION X X o ASYMMETRIC BEHAVIOR X X X X

PROPRIETARY sq DUKE POWER CO.

ALTERNATE SG NODALIZATIONS AND MODELING o OBJECTIVES

- IMPROVED PHASE SEPARATION

- ASSESS TUBE BUNDLE UNCOVERY IMPACT ON HEAT TRANSFER o EFFORTS

- ALTERNATE INITIALIZATION APPROACHES

- LOCAL CONDITIONS HEAT TRANSFER MODELING (2 NODE SG SECONDARY)

- 15 NODE SG SECONDARY MODEL

- 11 NODE SG SECONDARY MODEL o

SUMMARY

PROPRIETARY s DUKE POWER CO.

ALTERNATE INITIALIZATION APPROACHES o RETRAN DYNAMIC SLIP

- INITIALIZATION DIFFICULTIES

- "GRADUAL" SEPARATION o RETRAN SEPARATOR MODEL

- NO BETTER THAN BUBBLE RISE

- NO BASIS FOR PERFORMANCE

- NOT WIDELY USED o BROADER USE OF BUBBLE RISE MODEL

- STACKING PROBLEMS

- INCOMPATIBLE WITH RECIRCULATION

- SELECTION OF PARAMETER VALUES

PROPRIETARY s DUKE POWER CO.

2-NODE LOCAL CONDITIONS MODEL o "BOILING POT" MODEL WITH BUBBLE RISE o TUBE BUNDLE UNCOVERY HEAT TRANSFER DEGRADATION PREDICTED o CANNOT HANDLE SUPERHEATING o CANNOT HANDLE RECIRCULTION AND MASS FLUX EFFECTS o CANNOT REPRESENT NR SG LEVEL o APPLIED TO:

- LOSS OF MFW

- LOSS OF ALL FEEDWATER LESSONS LEARNED o HEAT TRANSFER DEGRADATION ADEQUATELY MODELED o SG MASS MUST APPROACH DRYOUT BEFORE HEAT TRANSFER IS AFFECTED

- POST-TRIP DECAY HEAT OF CONCERN

- 90% OF INVENTORY MUST BE BOILED OFF

'BEFORE ANY IMPACT o CANNOT BE USED "STAND-ALONE" o INSIGHTS GAINED BUT LIMITED APPLICABILITY

PROPRIETARY DUKE POWER CO.

Lu

0 ta CNS LOSS OF ALL FEEDWATER 5.00E+08 H

E 4.00Et08 A

106 107 105 108 104 109 R3.OOE+08 A103 + 110 tI E2.00E108 R

R A

E T~1 1.0E0 10

-0 00 E+00 Or0 o~ ~~ ~ ~ ~ ~~ E18000.~~r-~~

~~~1

~~--,-i ,-i-- ~0 0>

H 0 200 400 600 800 1000 1200 1400 160 10.2000 TIME (SECONDS) t0

PROPRIETARY DUKE POWER CO.

15-NODE SG MODEL o ADDITIONAL TUBE BUNDLE REGION DETAIL WITHIN THE FRAMEWORK OF THE STANDARD NODALIZATION o PSEUDO TWO-DIMENSIONAL MODEL o PHASE SEPARATION DRIVEN BY HEAT TRANSFER (HEM NODES) o FEW LIMITATIONS

- NR LEVEL SIMULATED

- RECIRCULATION MODELED

- SUPERHEAT POSSIBLE

- NO CROSSFLOW JUNCTIONS o APPLIED TO LOSS OF MFW LESSONS LEARNED o EXCESSIVE SUPERHEAT PREDICTED o ABSENCE OF CROSSFLOW IS MAJOR PROBLEM o CONFIRMED VALIDITY OF 7 NODE MODEL

PROPRIETIARY

  • - DUKE POWER CO.

LI)

-L - - -

l0

PROPRIETARY 1 DUKE POWER CO.

11-NODE SG MODEL o INTENDED TO AVOID CROSSFLOW PROBLEM THAT OCCURRED IN 15-NODE MODEL o ALL OTHER CAPABILITIES SIMILAR TO 15-NODE MODEL LESSONS LEARNED o INADEQUATE PHASE SEPARATION OBSERVED.

RECIRCULATION OF LIQUID CONTINUES BEYOND THE INVENTORY AT WHICH IT IS EXPECTED TO STOP o CONFIRMED VALIDITY OF 7-NODE MODEL

PROPRIETARY DUKE POWER CO.

SUMMARY

o ALTERNATE NODALIZATIONS HAVE BEEN DEVELOPED AND TESTED o INSIGHTS HAVE BEEN GAINED REGARDING THE IMPACT OF SG INVENTORY DEPLETION ON HEAT TRANSFER o DUE MAINLY TO PHASE SEPARATION MODELING LIMITATIONS, NO BETTER NODALIZATION APPROACH HAS BEEN ACHIEVED o FOR FSAR CHAPTER 15 TRANSIENTS IT IS EXPECTED THAT THE BASE MODEL NODALIZATION WILL REMAIN ADEQUATE

- EXTENSIVE SG INVENTORY DEPLETION IS NOT ENCOUNTERED

- A MATCH BETWEEN HEAT SOURCES AND AFW FLO WRATE IS TYPICALLY ESTABLISHED IN AN ACCEPTABLE TIME FRAME

- AFW REALIGNMENTS CAN BE CREDITED OUT IN TIME

PROPRIETARY DUKE POWER CO.

Reactor Vessel Upper Head Nodalization

  • The upper head region outside the control rod guide tubes and UHI support columns is modeled with a single RETRAN-02 node
  • The typical concern with respect to noding in this region is sufficiently accurate modeling of voiding in the reactor vessel upper head during depressurization transients
  • Voiding is less of a concern for the McGuire and Catawba reactor vessels since this region is at the cold leg temperature during normal operation.

This is accomplished by diverting 3.7% of the vessel inlet flow to the upper head.

PROPRIETARY DUKE POWER CO.

UPPER SUPPORT PLATE SPRAY NOZZLE 000 SUPPORT COLUMN-----

DOWNCOMER REGION -- GUIDE TUBE UPPER CORE PLATE Reactor Upper Head Region Flow Paths

PROPRIETARY DUKE POWER CO.

T ana 0*

UT/C LOCATIONS W. B. McGuire Unit 1 4-Loop Upper Head Thermoccuple Schematic

PROPRIETARY DUKE POWER CO.

Reactor Vessel Upper Head Nodalization

  • Not modeling an upper layer of hotter fluid within the upper head is conservative for DNB evaluations since voiding of this layer would retard depressurization
  • The relevant benchmark of a loss of offsite power at McGuire Unit 1 showed a sufficiently accurate prediction of upper head temperature. As discussed in the September 8, 1989 response to Question 11b, a more accurate prediction would be obtained by adjusting the loss coefficient of the reactor coolant pump as it coasts down, as shown on the following figure.

PROPRIETARY DUKE POWER CO.

Id IL ow 0

Reactor Vessel PROPRIETARY DUKE POWER CO.

Core Nodalization

  • Although three axial and one radial nodes are judged to be adequate for most transients, additional noding is used in transient-specific situations as necessary, e.g., two radial nodes for the steam line break analysis
  • The number of axial nodes was chosen to provide some capability to model axial power distribution and axial fuel and moderator temperature distributions to calculate reactivity feedback.
  • The axial power distribution among the three nodes is conservatively modeled where it is deemed important

Reactor Vessel PROPRIETARY DUKE POWER CO.

Core Nodalization

  • Extremely conservative axial and radial power distributions are modeled as necessary, e.g., ensuring that the RETRAN-02 axial steam line break axial distribution is more top-peaked than that of a detailed three-dimensional nuclear physics code
  • While the RETRAN-02 noding described on the previous slide is typical, the VIPRE-01 core thermal-hydraulics noding is always very detailed, i.e., 40 axial nodes and 14 channels gi

PROPRIETARY DUKE POWER CO.

Transient-Specific Deviations from Modeling in DPC-NE-3000

  • Local conditions heat transfer in the pressurizer wall heat conductors used to more realistically model physical phenomena:
  • Has been or will be used in general, with twelve wall conductor nodes rather than the one specified in DPC-NE-3000
  • Model is more important for medium to long duration transients
  • Model tends to somewhat mitigate the tendency of the RETRAN-02 pressurizer model to overpredict pressure during an insurge

PROPRIETARY DUKE POWER CO.

Transient-Specific Deviations from Modeling in DPC-NE-3000

  • Local conditions heat transfer in the pressurizer wall heat conductors:
  • Results are reviewed to ensure that this model gives a conservative response.

The steam line break analysis did not use this model because a sensitivity study indicated that this model gave a somewhat higher pressure result for that transient

  • Other transients for which DPC-NE-3002 inadvertently omitted a reference to this model will also be analyzed using it

PROPRIETARY DUKE POWER CO.

Transient-Specific Deviations from Modeling in DPC-NE-3000

  • Single node steam generator secondary used in uncontrolled bank withdrawal from subcritical since the duration of the transient is short and the steam generator secondary response is not important
  • A triple loop model is used in the feedwater line break analysis because of the differences in the boundary conditions among the broken loop, the intact loop not receiving auxiliary feedwater, and the two intact loops receiving auxiliary feedwater

Turbine Trip

  • Acceptance Criteria e Peak Secondary Pressure Case
  • Initial Conditions
  • Boundary Conditions
  • Results
  • Peak RCS Pressure Case
  • Initial Conditions
  • Boundary Conditions
  • Results
  • Nodalization Sensitivity
  • Summary

Turbine Trip Acceptance Criteria

  • Peak RCS pressure remains below 110%
  • of the design value
  • Peak Main Steam System pressure remains below 110% of the design value

Initial Conditions for Peak Secondary Pressure Analysis Duke Westinghouse NSSS Power 3498 MW 3425 MW Pzr Pressure 2310 psia 2250 psia Pzr Water Volume 1253 ft3 1175 ft3 RCS Flow 420,000 gpm 388,880 gpm RV Avg. Temp. 594.8 0F 590.1oF MFW Temperature 440 0 F 440 0 F

Boundary Conditions for Peak Secondary Pressure Analysis

  • RPS trip function is overtemperature AT (Duke) vs. high pressurizer pressure (Westinghouse)
  • Rod control in manual
  • No credit taken for steam dump or steam line PORVs
  • Turbine stop valves close in 0.02 seconds (Duke) vs. 0.1 seconds (Westinghouse)

Table 15.2.3-1 (Page 1 of 5)

Time Sequence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary Systems Accident Event Time (sec)

Turbine Trip

1. With pressurizer Turbine trip, loss 0.0 0.0 control (minimum of main feed flow reactivity feedback)

Initiation of steam 7.0 5.f release from steam gen erator safety valves High pressurizer pressure 7.3 13.0 reactor trip setpoint (or7-)

reached Rods begin to drop 9.3 /q.s Peak pressurizer 11.0 v/A pressure occurs Minimum DNBR occurs 11.5 Al/A

2. With pressurizer Turbine trip, loss 0.0 control (maximum of main feed flow reactivity feedback)

Peak pressurizer 7.0 pressure occurs Initiation of steam 7.0 release from steam gen erator safety valves Low-low steam gen- 62.6 erator level reactor trip setpoint reached Rods begin to drop 66.6 Minimum DNBR occurs (1)

(1) DNBR does not decrease below its initial value.

1986 Update

FSAR SECTION 15.2.3 TURBINE TRIP 120 MAXIMUM SECONDARY PRESSURE CASE 10 ................ . . ............... . ...... ..... ....... .......

0 o

w......- -- - . . . . . . - - .-

0 40 - ....

20 0 10 ?3 40 TIME (SECONDS)

LOOP AVERAGE TEMPERATURE (DEG F) 0 0LA0 0 ___

100 rnn c')

FSAR SECTION 15.2.3 TURBINE TRIP 100 MAXIMUM SECONDARY PRESSURE CASE 90 - -

80 -

Lu C,)

050 3) 23304 0 0 20 30 40 TIME (SECONDS)

O

FSAR SECTION 15.2..3 TURBINE TRIP 2700 MAXIMUM SECONDARY PRESSURE CASE 2600 2500 0.00 2400 cl)

Ul) 0C Lu cl) 2200 CL) 2100 2000 - ... ... ----

1900 0 10 20 30 40 TIME (SECONDS)

Initial Conditions for Peak RCS Pressure Analysis Duke Westinghouse NSSS Power 3498 MW 3425 MW Pzr Pressure 2280 psia 2250 psia Pzr Water Volume 1253 ft3 1175 ft3 RCS Flow 373,596 gpm 388,880 gpm RV Avg. Temp. 594.8 0F 590.1 0 F MFW Temperature 440oF 440oF

Boundary Conditions for Peak RCS Pressure Analysis

  • RPS trip function is high pressurizer pressure
  • Rod control in manual
  • No credit taken for steam dump or steam line PORVs
  • Turbine stop valves close in 0.02 seconds (Duke) vs. 0.1 seconds (Westinghouse)
  • No credit taken for pressurizer spray and PORVs

Table 15.2.3-1 (Page 2 of 5)

Time Sequence Of Events For Incidents Which Cause A Decrease In Heat Removal By The Secondary System Accident Event Time (Sec)

Turbine Trip

3. Without pressurizer Turbine trip, loss 0.0 0.0 control (minimum of main feed flow reactivity feedback)

High pressurizer pressure 3.7 1.5 reactor trip setpoint reached Rods begin to drop 5.7 &.5 Initiation of steam 6.5 release from steam gen erator safety valves Peak pressurizer 7.0 7.7 pressure occurs Minimum DNBR occurs (1) AIIA

4. Without pressurizer Turbine trip, loss 0.0 control (maximum of main feed flow reactivity feedback)

High pressurizer pressure 3.7 reactor trip setpoint reached Rods begin to drop 5.7 Initiation of steam 6.5 release from steam gen erator safety valves Peak pressurizer pressure occurs 6.5 1

Minimum DNBR occurs (1)

(1) DNBR does not decrease below its initial value.

1986 Update

FSAR SECTION 15.2.3 TURBINE TRIP 2700 MAXIMUM PRIMARY PRESSURE CASE 2600 2500 1 a

CL 2300 --- fT cc w 00 200 CL' 2100 c10 200 33 4 2000------

1900, 0 10 20 30 40 TIME (SECONDS)

PROPRIETARY DUKE POWER CO.

Turbine Trip Nodalization Sensitivity

  • Turbine trip peak secondary pressure case was reanalyzed with 11 rather than 2 steam line nodes.
  • The 11 node result differs from the submitted result by less than 1 psi

Turbine Trip Summary

  • Peak RCS pressure result is 2689 psia (Duke) vs. approximately 2570 psia (Westinghouse case with sprays and PORVs off and "minimum" feedback)
  • Peak secondary pressure result is 1308 psia (Duke). Westinghouse does not quote a result or present a plot
  • Westinghouse models the pressurizer safety valves as full open at 3% above the set pressure (2575 psia)
  • Duke models the pressurizer safety valves as full open at 3% above the drifted set pressure (2652 psia)
  • All acceptance criteria are met

Locked Rotor

  • Acceptance Criteria
  • Core Cooling Capability Case
  • Initial Conditions
  • Boundary Conditions
  • Results
  • Peak RCS Pressure Case e Initial Conditions
  • Boundary Conditions
  • Results
  • Nodalization Sensitivity
  • Summary

Acceptance Criteria

  • Offsite doses do not exceed a small fraction (10%) of the 10 CFR Part 100 guidelines
  • Two separate analyses are required to conservatively evaluate the transient with respect to both acceptance criterion

0 Initial Conditions for Core.

Cooling Capability Analysis Duke Westinghouse NSSS Power 3498 MW 3493 MW Pzr Pressure 2208 psia 2280 psia Pzr Water Volume 930 ft3 1092 ft3 RCS Flow 373,596 gpm 377,000 gpm RV Avg Temp. 594.8 0F 594.1 0F MFW temperature 440'F 440 0 F

Boundary Conditions for Core Cooling Capability Analysis

  • RCCA drop time of 2.2 seconds (Duke) vs 2.7 seconds (Westinghouse)
  • Rod control in manual (both)

e No credit taken for steam dump or steam line PORVs (both)

  • Penalty taken for operation of pressurizer spray and PORVs (Duke)

Sequence of Events for Core Cooling Capability Analysis Rotor on one pump locks 10.0 1.0 Reactor Coolant Pump Shaft Seizure (Locked Rotor)

Low flow reactor trip 10.05 1.05 setpoint reached Rods begin to drop 12.05 2*CE Maximum RCS pressure Li5.2 occurs Maximum clad 41.5 2.2 temperature occurs

FSAR SECTION 15.3.3 LOCKED ROTOR 1.2 faulted loop 1.0 ............... intact loop 0_ __ __ -: in------------- ------- -

00 . . . . . . ...................................

LL .................

~ ~

0 . ... . ....................

0.4~~~~~~~~

....... ~ ~

..... ~ ~. . ............

z 0

o U

0.2

-0.4 2 3 9 TIME (SECONDS)

FSAR SECTION 15.3.3 - LOCKED ROTOR 120 z 80_ _ __ _ __ _ _ _

0 z

U-U CLD 40 -- --- -------- ........... -. - - -.----- ..... ....... ... ............................. ............. *... ......

w 0.0 0 32 4 9 1 TIE SE-NS

FSAR SECTION 15.3.3 LOCKED ROTOR 1.1 1.0 z

0 O0.9 z 0.8 ........... . ..... .............

IL 0

z 0 0.7 LL 0.

Li 0.1 0.0 0(O TIME (SECONDS)

FSAR SECTION 15.3.3 - LOCKED ROTOR 620 ..................... . ............... .... . . .. . . . ............ ..... faulted loop ................

.............. intact loops 0 600 w

cc ..................... ............. .... . . .............

.5 0... .. .. ~... _ _. --. ---..--------- - _ _

LU 0

w580 ---- _ -__ _ -- . _ _ _ _ _------- - - ._ _ -----------

TIM (SCODS

FSAR SECTION 15.3.3 - LOCKED ROTOR 500 U)

Li ............

cnI m 2300 .......---..-

CO U0 N

C 2200 L

2100 ..---.

2000 0 1 2 3 4 5 6 7 8 9 10 TIME (SECONDS) 04

Initial Conditions for Peak RCS Pressure Analysis Duke Westinghouse NSSS Power 3498 MW 3493 MW Pzr Pressure 2310 psia 2280 psia Pzr Water Volume 1253 ft3 1092 ft3 0

RCS Flow 373,596 gpm 377,000 gpm RV Avg Temp. 594.8 0F 594.1 0F MFW temperature 440'F 440oF

Boundary Conditions for Peak RCS Pressure Analysis

  • RCCA drop time of 2.2 seconds (Duke) vs 2.7 seconds (Westinghouse)
  • Rod control in manual (both)
  • No credit taken for steam dump or steam

101

  • No credit taken for pressurizer spray and PORVs (both)

107, Sequence of Events for Peak RCS Pressure Analysis Reactor Coolant Pump Rotor on one pump locks . .0 Shaft Seizure (Locked Rotor)

Low flow reactor trip 10.05 1.05 setpoint reached Rods begin to drop Z.Z.05 2.OS Maximum RCS pressure AI..2 occurs Maximum clad (3.5 -

temperature occurs

FSAR SECTION 15.3.3 LOCKED ROTOR 2600 2500 CC C 2400 J____

U)

Cl) 2300 2200 I I

.4 10 TIME (SECONDS)

PROPRIETARY DUKE POWER CO.

Nodalization Sensitivity

  • Number of axial core nodes doubled, from 3 to 6

Summary

  • Peak RCS pressure result is 2558 psia (Duke) vs 2613 psia (Westinghouse)
  • Westinghouse assumes an MTC of +7.0 pcm/F, while Duke applies Technical Specification limits.
  • Westinghouse models the pressurizer safety valves as full open at 3% above the set pressure (2575 psia)
  • Duke models the pressurizer safety valves as full open at 3% above the drifted set pressure (2652 psia)
  • All acceptance criteria are met

Locked Rotor VIPRE DNBR Analysis NRC I DUKE Meeting October 7&8, 1991

Locked Rotor VIPRE DNBR Analysis PROPRIETARY DUKE POWER CO.

  • The reactor core remains symmetric in neutron power as well as coolant properties
  • The 14 channel model with the fuel conduction model is used to perform the transient DNBR and fuel rod cladding temperature calculations
  • The fuel conduction model is identical to that presented in DPC-NE-3001 except that the gas gap conductivity is maintained constant at its initial value throughout the transient

.4

lot PROPRIETARY DUKE- POWER CO.

14 Channel VIPP! Model with Fuel Conduction Model 141 13 Conducting Fuel Rod Instrumentation Control Rod Guide Tube Guide Tube 0

Locked Rotor VIPRE DNBR Analysis (Cont'd)

  • Transient boundary forcing functions obtained from RETRAN analysis are:
  • Core average neutron power
  • Core exit pressure
  • Core inlet flow rate
  • Core inlet temperature

Locked Rotor VIPRE DNBR Analysis (Cont'd)

  • The BWCMV critical heat flux (CHF) correlation is utilized for the DNBR calculation
  • The Non-Statistical Core Design (NSCD) methodology is utilized
  • The DNBR limit is 1.331 (1.331 = 1.1 x 1.21; where 1.21 is the correlation design limit and 1.10 adds 10% margin)

Locked Rotor VIPRE DNBR Analysis (Cont'd)

  • Methodology
VIPRE calculates the transient pin cladding temperature and maximum allowable radial peaking (MARP) during the transient such that DNB would not occur
  • The MARP result is used to determine the amount of fuel pin in the core having DNB (failed fuel)

Locked Rotor VIPRE DNBR Analysis (Cont'd)

PROPRIETARY DUKE POWER CO.

  • MARP curves
  • A spectrum of axial power distribution is analyzed. The magnitude and shape of the axial profile is varied to cover the full range of profiles resulting from the nuclear design analysis
  • For each given axial peak value, utilizing the RETRAN transient boundary conditions, the pin radial peak is searched until the DNBR limit of 1.331 is obtained
  • The axial and radial peaks are assumed to be constant throughout the transient

Locked Rotor Accident Maximum Allowable Radial Peaks for 2.2 Second Control Rod Iosertion Time MBW Fuedl Type 1.5 X1.3 2 AP = 1.3 IAP = 1.5

- -- AP = 1.7 C 0.5 0.6 0.7 0.8 0.9 Normalized Axial Peak Location

FSAR SECTION 15.3.3 LOCKED ROTOR 760 LU LI IL w TE LUU (740 .

740 0 12 3 4 TIME (SECONDS)

Locked Rotor VIPRE DNBR Analysis (Cont'd)

  • Results The maximum clad inner surface temperature is 758.5 OF during the transient Utilizing the MARP curves, nuclear design performs the fuel pin census to determine the amount of fuel pin experiencing DNB (failed rod)

Core will remain in place and intact with no loss of core cooling capability since the clad surface temperature is far less than the clad deformation temperature

Locked Rotor VIPRE DNBR Analysis (Cont'd)

  • Differences between the original FSAR and DPC analysis methods PROPRIETARY DUKE POWER CO.

original FSAR DPC

- 20% neutron power - No such neutron power increase during the increase during the transient transient

- DNB is assumed to occur - No such assumption is in the beginning of the made. DNB is explicitly transient calculated The gap conductivity is - The gap conductivity is forced to change from its held constant at its initial initial value to 10,000 value throughout the Btu/hr-ft2-F transient

  • *e Uncontrolled Bank Withdrawal at Power
  • Acceptance Criteria
  • Core Cooling Capability Analysis
  • Initial Conditions
  • Boundary Conditions
  • Results
  • Peak RCS Pressure Analysis
  • Initial Conditions
  • Boundary Conditions
  • Results
  • Summary

Transient Ov&view

  • Bank withdrawal
  • Changes power distribution
  • Inserts positive reactivity
  • Positive reactivity increases neutron power
  • Neutron power increases core heat flux
  • Heat flux increases RCS temperature
  • RCS temperature
  • Increases RCS pressure
  • Reactor protection provided by
  • Overtemperature Delta T (OTDT)
  • Power range high flux (high setpoint)
  • High pressurizer pressure
  • DNBR analysis sensitivity studies
  • Initial power
  • Withdrawal rate
  • Reactivity feedback
  • Pressurizer pressure control-

Acceptance Criteria

. Two acceptance criteria

  • DNB is avoided
  • Peak RCS pressure less than 110% of design (2750 psia)
  • Duke performs a separate analysis for each acceptance criterion
  • DNB analysis employs SCD methodology
  • Allowances in peak RCS pressure analysis bound uncertainties
  • Westinghouse performs only one analysis
  • Analysis evaluates DNB using ITDP methodology
  • Peak RCS pressure not specifically addressed
  • 0 Initial Conditions for Core Cooling Capability Analysis (10% Power)

Parameter Duke Westinghouse NSSS Power, MW 361 343 Pzr Pressure, psia 2250 2250 Pzr Water Volume, ft3 408 544 RCS Flow, gpm 382,000 388,880 RV Avg. Temp., oF 560.4 560.2 MFW Temperature, 'F 307 241 O

Initial Conditions for Core Cooling Capability Analysis (100% Power)

Parameter Duke Westinghouse NSSS Power, MW 3431 3425 Pzr Pressure, psia 2250 2250 Pzr Water Volume, ft3 930 1092 RCS Flow, gpm 382,000 388,880 RV Avg. Temp., oF 590.8 590.1 MFW Temperature, oF 440 440

  • e Boundary Conditions for Core Cooling Capability Analysis
  • Pressurizer safety valves
  • Duke: Valves modeled to minimize RCS pressure, full open at

-2% setpoint (2450 psia).

  • Westinghouse: Valves full open at +3% setpoint (2575 psia).
  • Steam line safety valves
  • Duke: Valves maximize SM pressure, full open at +3% of +3% drifted high setpoint. Bank 5 lift at 1282 psia, full open at 1320 psia.
  • Westinghouse: Valves full open at +3% setpoint. Bank 5 full open at 1277 psia.
  • No credit taken for steam line PORVs in order to minimize heat transfer
  • Turbine in load control mode to maximize power mismatch
  • 0 Reactor Trip Setpoints for Core Cooling Capability Analysis
  • Duke: 2445 psia value bounds uncertainties and includes additional margin.
  • Westinghouse: 2425 psia
  • Duke: 113.2 % RTP value bounds uncertainties and includes additional margin. Equivalent setpoint includes effects of power dependent rod shadowing Power, % RTP Eq. setpoint, % RTP 100 113.2 50 117.1 10 128.6
  • Westinghouse: 118 % RTP.

Sensitivity Stutes for Core Cooling Capability Analysis

  • Three initial core power levels
  • Duke: 10%, 50%, 100%
  • Westinghouse: 10%, 60%, 100%
  • Spectrum of withdrawal rates
  • Duke: maximum 65 pcm/sec
  • Westinghouse: maximum 110 pcm/sec
  • Two reactivity feedback conditions
  • Duke: Coefficients for most positive feedback at both BOC and EOC
  • Westinghouse: Coefficients for both most positive feedback and most negative feedback
  • Pressurizer Pressure Control; high pressure is benefit in DNBR calculation, but delays OTDT reactor trip
  • Westinghouse: No pressure control sensitivity

-I:

Results for Core Cooling Capability Analysis Most limiting case

  • Duke: 10% initial power, BOC, 15 pcm/sec, sprays/PORVs operable Calculated MDNBR = 1.55, Limit Value = 1.55
  • Westinghouse: 60% initial power, minimum feedback, 6 pcm/sec Calculated MDNBR = 1.7, Limit Value = 1.49 Sequence of events for most limiting case Event Time, seconds Initiate Bank withdrawal 0.0 Pressurizer sprays full on 19.4 Pressurizer PORVs full open 21.8 OTDT setpoint reached 37.2 Control rod insertion begins 38.7 MDNBR occurs 40.3 U'

FSAR SECTION 15.4.2 UCBW @ 10% POWER 2.65

$I 4I 2.45 - - - - ca e Ge La 2.25

OTDT TRIP HP TRIP HF TRIP 4F T 2.05 onh 1.85 42 1.65 14 1.510 10 RATE, PCM / SEC

FSAR SECTION 15.4.2 UCBW @ 100% POWER 2.45

,2.25 0 ____ pj~~ - HF TRIP 2.05 0-6 1.85 1.65 14f e 1.45 100 1 10 RATE, PCM i SEC

POWER 10 FSAR SECTION 15.4.2 UCBW 1.2 1.0 ................... . .......... ................

01 z

LL ..........

0.0 ....

00 00 04 LU 04 0.2 02.

20 4060.

TIME (SECONDS) CI

FSAR SECTION 15.4.2 IJCI3W I()0%POWER 120 ~1~

100. ... ................. ...... .

4 ~. ..-......... . .........-.. . . . . . . ...

Z80 ..................... .Oi T

0 z

0

'0 U63 - - .. . . ~ . .- - - . ..... ..- -------

00 0

z 0 ...................................................

cc 34 ..................

20 60 80 040 T E SECONDS)

FSAR SECTION 15.4.2 UJCBW (y@ 10 % POWER 600 LZ 590 0L L

00 cc LJU LI 560 60 80 20 40 0

TIME (SECONDS)

FSAR SECTION 15.4.2 UCIBW @ 1() % POWER 2600 *

.240 . . . . .. ... .... -.

I 2200 ...................

220 4 cc CL U 2000 ......

U)

U)

L 1600 40 60 80 0 20 JIME (SECONDS)

Initial Conditions for Peak RCS Pressure Analysis Parameter Duke Westinghouse NSSS Power, MW 361 3425 Pzr Pressure, psia 2250 2250 Pzr Water Volume, ft3 709 1092 RCS Flow, gpm 382,000 388,880 RV Avg. Temp., oF 560.4 590.1 MFW Temperature, F 307k 440

Boundary Conditions for Peak RCS Pressure Analysis

  • High pressure reactor trip setpoint of 2430 psia bounds initial condition uncertainties and includes additional margin. Westinghouse value for DNB analysis is 2425 psia.
  • No credit for pressurizer sprays and PORVs
  • Pressurizer safety valves
  • Duke: Valves maximize RCS pressure. Lift at +3% setpoint (2575 psia),

full open at +3% accumulation ( 2652 psia).

  • Westinghouse: Valves full open at +3% setpoint (2575 psia).
  • Steam line safety valves
  • Duke: Valves maximize SM pressure, full open at +3% of +3% drifted high setpoint. Bank 5 lift at 1282 psia, full open at 1320 psia.
  • Westinghouse: Valves full open at +3% setpoint. Bank 5 full open at 1277 psia.

Results for Peak RCS Pressure Analysis Peak pressure of 2738 psia, 12 psi margin to acceptance criterion Sequence of events Event Time, seconds Initiate Bank Withdrawal 0.0 Reactor trip setpoint reached 12.6 Pressurizer Safety Valves Lift 14.5 Control Rod Insertion Begins 14.6 Peak Pressure Occurs 15.8

FSAR SECTION 15.4.2 UCBW @ 10% POWER 2750 00 2700.00 2650.00 2600.00 9 2550.00 MAXIMUM PRESSURE CASE 2500.00 2450.00 2400.00 2350.00 --

2300.00 10.00 12.00 14.00 16.00 18.00 20.00 0.00 2.00 4.00 6.00 8.00 TIME (SECONDS)

  • 0 0 Summary
  • Main differences Duke and Westinghouse analyses:
  • Rod shadowing
  • Trip Setpoints
  • Pressure control sensitivity

. Feedback sensitivity

  • Peak pressure analysis
  • Acceptable results for both core cooling capability and peak pressure analyses

Uncontrolled-Bank Withdrawal at Power VIPRE DNBR Analysis NRC / DUKE Meeting October 7&8, 1991

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis PROPRIETARY DUKE POWER CO.

  • The reactor core remains symmetric in neutron power as well as coolant properties
  • The 14 channel model without the conduction model described in DPC-NE-3000 is used to perform the transient DNBR calculation

PROPRIETARY rflVT "

14 Channel VIFRZ Model 1 2 3 5 /

5 6 7 9 9 10 11 12 13 10 9 10 11 12 16 17 1 1319 14 20 Fuel Rod Instrumentation Control Rod Guide Tube Guide Tube 0 1

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd)

  • Transient boundary forcing functions obtained from RETRAN analysis are:
  • Core average heat flux
  • Core exit pressure
  • -Core inlet flow rate
  • Core inlet temperature Ic

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd)

  • Uncontrolled bank withdrawal (UCBW) at 10%, 50%, and 100% FP cases are analyzed
  • The BWCMV critical heat flux (CHF) correlation and the Statistical Core Design (SCD) methodology are utilized for the DNBR calculation
  • The DNBR limit is 1.55 (SCD correlation limit of 1.40 plus margin)

Uncontrolled Bank Withdrawal at Power PROPRIETARY VIPRE DNBR Analysis (Cont'd) PIIKF PONFR CO.

  • Methodology The relationship between the reactivity insertion rate and the lowest minimum DNBR (MDNBR) during the transient is first determined. For each power level case, there is a reactivity insertion rate that would result in the most limiting MDNBR during the accident The statepoint condition of this most limiting MDNBR point is utilized to generate the post accident maximum allowable radial peaking (MARP) curves for each power level case

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd) PROPRIETARY DUKE POWER CO.

  • What is MARP?
  • A spectrum of axial power distributions is analyzed. The magnitude and shape of the axial profile are varied to cover the full range of profiles resulting from the nuclear design analysis For each given axial power distribution, utilizing the statepoint boundary conditions of the most limiting MDNBR point, the pin radial peak is searched until the BWCMV CHF DNBR limit of 1.55 is obtained

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd) PROPRIETARY DUKE POWER CO.

  • UCBW at 100% FP Power distribution Axial peak = 1.55 at x/l = 0.7 Radial peak = 1.50 Axial and radial power distributions are assumed to be constant throughout the transient Other axial and radial power distributions are also used to ensure that the time at which the most limiting MDNBR occurs during the transient is independent of the power distribution utilized Reactivity insertion rate versus MDNBR curve MARP curves

FSAR SECTION 15.4.2 UCBW @ 100% POWER 2.45 2.25 HF TRIP 2.05 HP TRIP 1.85 1.65 1.45 10 100 RATE, PCM / SEC

PROPRIETARY DUKE POWER CO.

Maximum Allowable Radial Peaks for Uncontrolled RCCA Bank Withdrawal at 100% FP with Positive Feedback (MBW Fuel - 2.2 Second Scram Time) 2.0 a1.8 Ce 1.6 E - e - AP =1.1 E AP =1.3 2 " AP =1.5 1.4

-AP = 1.7 o .-

AP =UO1.9

--0 -- AP =2.1 1.2 .

0.0 0.2 0.4 0.6 0.8 1.0 Normalized Axial Peak Location

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd) PROPRIETARY DUKE POWER CO.

  • UCBW at 50% FP Power distribution Axial peak = 1.55 at x/l = 0.7 Radial peak = 1.725 The axial and radial peakings are assumed to be constant throughout the transient Reactivity insertion rate versus MDNBR curve MARP curves

FSAR SECTION 15.4.2 UCBW @ 50 % POWER 2.65 2.45

- HF TRIP 2.25 -------- OTDT TRIP 0

2.05 1.85 1.65 1.45 RATE, PCM / SEC 100

  • eg PROPRIETARY Maximum Allowable Radial Peaks for Uncontrolled DUKE POWER CO.

RCCA Bank Withdrawal at 50% FP with Negative Feedback (MBW Fuel - 2.2 Second Scram Time) 2.0 1.8 0

o 1.6 E -- e - AP = 1.1

. AP =1.

1.4 - -o ~ AP =1.7

.. AP = 1.9

--o--AP = 2.1 1.2.

0.0 0.2 0.4 0.6 0.8 1.0 Normalized Axial Peak Location

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd) PROPRIETARY DUKE POWER CO.

  • UCBW at 10% FP Power distribution Axial peaking = 1.55 at x/l = 0.7 Radial peaking = 1.725 The axial and radial peakings are assumed to be constpnt throughout the transient Reactivity insertion rate versus MDNBR Curve MARP curves

FSAR SECTION 15.4.2 UCBW @ 10% POWER

.2.65 iif 2.45 I if 2.25

OTDT TRIP HP TRIP

.................... HF TRIP 2.05 4

1.85 1.65 1.45 1 10 100 RATE, PCM / SEC

PROPRIETARY Maximum Allowable Radial Peaks for Uncontrolled DUKE POWERYCO.

RCCA Bank Withdrawal at 10% FP with Positive Feedback (MBW Fuel - 2 2 Second Scram Time) 2.0 0

0 6P - --. = .

EI AP = 1.3 E .AP = 1.5 Mi1.4 - -- +--AP = 1.7 AP = 1.7

-P-2.1AP = 2.1 aW 1.2 0.0 0.2 0.4 0.6 -0.8 1.0 Normalized Axial Peak Location

Uncontrolled Bank Withdrawal at Power VIPRE DNBR Analysis (Cont'd)

  • Result
Utilizing the MARP curves, nuclear design performs the DNB margin check. The check results in positive DNB margin for all power level cases (no failed fuel)

PROPRIETARY

  • 113.2%Power Range High Flux Trip Setpoint Calculation
  • Process measurement accuracy (PMA) term of 1.7%for power calorimetric calculation unchanged from original McGuire licensing submittal
  • Rack terms of 0.5% (calibration accuracy, RCA), 0.25% (comparator setting accuracy, RCSA), 1.00% (drift, RD), and 0.50% (temperature effects, RTE) all unchanged from original McGuire licensing submittal
  • Channel statistical allowance (CSA) is the square root of [PMA 2 + (RCA + RCSA +

RD) 2 + RTE2] = 2.49% span or 2.99%RTP

  • Margin of 1.21%RTP was included to avoid a decrease in Technical Specification allowable value. Technical Specification setpoint of 109% + 2.99 +1.21 =

113.2%RTP

PROPRIETARY 113.2%Power Range High Flux Trip Setpoint Calculation

  • Original McGuire licensing submittal also used another PMA term for "neutron flux-water density in downcomer, shielding effects, detector placement" of 4.2% span (5%RTP). Duke accounts for this in the shadowing adjustment to get excore detector flux from point kinetics core average flux
  • Original McGuire licensing submittal, including this additional PMA term, had a CSA of 4.9% span and a margin of 2.6%

span. This gives a total allowance of 7.5%

span or 9%RTP. Technical Specification setpoint of 109% + 9% gives the Westinghouse safety analysis setpoint of

  • 118%RTP.

Minimum Technical Specification Indicated Pressurizer Pressure

  • If all four channels of pressurizer pressure control board indications are functioning at McGuire, indicated pressure is allowed to be only 8.5 psi less than the nominal 2235 psig value
  • If only three channels of pressurizer pressure control board indications are functioning at McGuire, indicated pressure is allowed to be only 5.2 psi less than the nominal 2235 psig value 0
  • 0 Pressure Uncertainty 1 2 3 4 Initial Indicated, psia 2250 2250 2250 2220 Initial Actual, psia 2220 2280 2250 2250 Trip Setpoint, psia 2400 2400 2430 2400 Initial Margin, psi 150 150 180 180 Actual Pressure at Trip, psia 2370 2430 2430 2430
  • For peak pressure analysis, col 2 is more conservative than col 1.
  • To avoid re-initialization, col 3 is used instead of col 2, preserving actual pressure at trip.
  • Col 4 is equivalent to col 3, only placement of uncertainty is different.

IMPACT OF TUBE BUNDLE UNCOVERY ON HEAT TRANSFER

I59 PROPRIETARY DUKE POWER CO.

K

-pop

CNS LOSS OF ALL FEEDWATER 5.00E+08 H

E 4.00E+08 A

T I107 106 108 1105 104 109 R 3.00E+08 A 103 I110 A

N S

F E 2.00E+08 R

R A

T 1.00E+08 E 1508 0 OOE+OO 200 400 600 800 1000 1200 1400 1600 1800. 2000 0

TIME (SECONDS) 10

CNS LOSS OF ALL FEEDWATER 100000 90000 S

G 80000 w

A 70000 T

E 60000 R

50000 M

A 40000 S

S 30000 20000 b

m 10000 0- I I I I I I I1IIIO 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 r TIME (SECONDS) 0>

P~

CNS LOSS OF ALL FEEDWATER 1.10E+03 S 1.08E+03 G

p 1.06E+03 R

E S 1.04E+03-S U

R 1.02E+03 E

P 1.00E+03 s

9 9.80E+02 cO 9.60 +02 -I I I I I I I I u 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 :E TIME (SECONDS)

CNS LOSS OF ALL FEEDWATER 5.63E+02 5.62E+02 5.62E+02 T 5.61 E+02 C 5.61 E+02 0

d 5.60E+02 5.60E+02 5.59E+02 5.59E+02 I nI 0

0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 ::

TIME (SECONDS) 0>

CNS LOSS OF ALL FEEDWATER 100 P 90 R

E 80 S

S U 70 R

60 z

E 50 R

40 L

E 30 V

E 20 L

% 10 0

0 200 400 600 800 1000 1200 1400 1600. 1800 2000 2200  :

TIME (SECONDS) 0r

0.00 L-I 2408.

f-4 2000 .

CIL CL 0.oo 1800.

100aAGl

  • 0~~~OG. ______

It a00.i r 1400 .

> 120C.

af m 600.

09 400.

200.

0.

loo IAO 102 103 t04 71"E f SEC I LOSS OF NORMAL FEEDWATER

!~MCGUIRE NUCLEAR STATION Figure 15.2.7-1 3/90

C.

Sao S40.

I oc, .0 ILt 1'.

so C

1200.

(3/9

MNS -LOSS OF MAIN FEEDWATER (A4107)

R 8 x

P 06 E

R 48 28 8 20e 486 see 886 l886 1206 TIME (SECONDS)

MNS - LOSS OF MAIN FEEDWATER (A4107) 2600 2see p

z R

2480 P

R E 2380 S

S U

R 20 E

P 2188 s

6 2888 29000 0 200 400 600 888 lo0 1200 TIME (SECONDS)

z 76 E

400 0 2w ________ 4_0_6wGoo_10_0_1_

TIE(ECNS 1-i xR

MNS -LOSS OF MAIN FEEDWATER (A4107)

S T

E A

m L

N E

R le E

S E 16o8 S

I e 280 40 68e a88 1688 1280 TIME (SECONDVS)

MNS -LOSS OF MAIN FEEDWATER (A4 107)

S N

R 6 L so E

v40 E

L x

28 0 208 408 6w8 a88 Iam 1200 TIME (SECONDS)

MNS - LOSS OF MAIN FEEDWATER (A4107) 640 638 620 T

H 610 0

TL 6000 D

vi E

F 6580 570 a 208 480 68 80 1000 1280 TIME (SECONDS) tb-

MNS -LOSS OF MAIN FEEDWATER (A4 107) 6w8 T

c L

D D(

E 578 6

568 0260 480 600 8am 166 1280 TIME (SECONDS) 0v

MNS LOSS OF MAIN FEEDWATER .(A4 107)

S G

iem E

0 esee N

D A

R 68888Q A 40000 S

28-- -- - -- -- --

8200 408 880 Boo 1808 1200 TIME (SECONDS)

LIQUID MAW --- TOTAL MASS - - - -

MNS - LOSS OF MAIN FEEDWATER (A4107) 1.1 S 6*9 0 8.7 I

D VOID FRACTIONS F 0.5 F TIME TM R

VOLUME 0 1200 C 152 0.0 0.77 T 8.3 153 0.67 0.77 154 0.89 0.89 0 155 0.93 0.94 N 156 0.28 0.99 157 0.0 0.82 S 158 0.99 1.00 O 200 400 600 88 l0 1200 TIME (SECONOS)

MNS - LOSS OF MAIN FEEDWATER (A4107)

S 14000 6

C CONDUCTOR 0 o 2i0 HEAT TRANSFER N COEFFICIENT D

D TIME u

- / 0 1200 C

T - 103 10077 973 o 9000- 0 104 9924 890 105 8658 809 R a-106 7786 753 107 7264 721 108 6567 681 T G60 -------

109 5722 630 C .... 110 1163 544 B

T 8 8 29 49 60 Be 180 120 148 168 18 200 TIME (SECONDS)

  • 0e G

s MNS - LOSS OF MAIN FEEDWATER (A4107)

C 1296. ..... .....

0 N

D CONDUCTOR U -HEAT TRANSFER U low TIME T # 0 1200 0 800 _-_-

103 9.92E5 2.25E4 104 5.24E5 1.30E4 --------- ----

105 4.27E5 1.13E4 P

106 1.83E5 5.12E3 -

0 108 2.81E5 8.71E3 -- - --

w 600 109 2.29E5 7.76E3 E 110 5.09E5 9.42E3 R

B 40 U

E a 20 48 60Be10 120 140 16O 100 2ee TIME (SECONDS)

MNS - LOSS OF MAIN FEEDWATER (A4102)

S 6,

SS E 866 C

0 N

D 60000 A

R y

40000 . - -

A S

0 200 40 600 800 1000 1200 1488 1600 1800 rn TIME (SECONDS)

LIQUID MASS - - - - - TOTAL MASS

MNS - LOSS OF MAIN FEEDWATER (A4102)

C 0

N 29 D

U C

T 0

R 15 P \

o -

E 10-7 R

co 0

T - ----

Z __20_0060 000__0 _20 140_100_80 E

z 02064080amlw10 10 G N E

R TIME (SECONDS) 0 183 184 185 186 167 lee 109 118--

MNS - LOSS OF MAIN FEEDWATER (A4102) 1300 S

T E

A 1288 L

N E

P 1100 C R

E S

S U

R E 8 P

S 8 200 400 600 080 188 1288 1480 1688 188 TIME (SECONDS)

Of"

MNS - LOSS OF MAIN FEEDWATER (A4102) 60 590

0. se 009 E 570 6

F 568 550 0 280 400 680 000 1000 1280 1400 1688 1888 TIME (SECONDS) O 0o

MNS -LOSS OF MAIN FEEDWATER (A4 102) z IR__ _ _ ___ _ _ _ __ _ _ _

70 L

E v .

68 48 8 200 480 666 888 18a8 1200 1488 1683 188w TIME (SECOND)S)

-4o~

FSAR SECTION 15.2.8 - FEEDLINE BREAK 90000 80000 - __ __

70000 - __ __

6 0000 - _ ___

c 00 _ __ _ _ _______ _ _ _ __ _____

wQ Cl AULT LOOP0 Q-D 30 0 200 400 600 800 1000 1200 1400 1600 1800 2000 TIME (SECONDS)

FSAR SECTION 15.2.8 - FEEDLINE BREAK 90000 -_____ __________

80000 -_____ _____ _____

INTACT SINGL ELOOP 70000 -_____

a. 60000 -_____________________

S50000 - _______ ___

C:) INTACT DOUBL= LOOP E 40000 - _____ __ ___

S30000 -_____ _____ _____

20000 - __ __

FAULTED LOOP 10000 -

0 ' 200 400 600 800 1000 1200' 1400 1600 1800 2000 mn TIME (SECONDS) 0

E
ox

I),) ak Po e (,olnpan. Al.S Tt (/o*IIf Nnclvar P'roduction IDept. I ire Prcsidfil P 0Bo 0,,v Nu).Iicr )pciulin lTo)33i75

(;1iorhu. A' .( 2S&'I -10 DUKE POWER October 16,1991 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 Numbers 50-269, -270, and -287 Handouts Presented in October 7 and 8, 1991 Meeting with NRC Staff and Contract Reviewers

References:

1) Letter, 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.

(Reaffirmation of Proprietary Affidavit for DPC-NE 2004)

5) Letter, M. S. Tuckman to NRC, September 25, 1991.

(Reaffirmation of Proprietary Affidavit for DPC-NE 3000)

On October 7 and 8, 1991, representatives of Duke Power met with NRC Staff and contract reviewers to discuss outstanding issues associated with three Topical Reports (References 1, 2, and 3),

which are currently undergoing review. Due to the large volume of information which was provided via handouts and slides, Duke has been requested to docket the presentation.

Accordingly, attached are a set of the handouts which were presented in the meeting. Please note that some of the information is identified as proprietary, and should be withheld from public disclosure pursuant to 10 CFR2.790. Affidavits attesting to the proprietary nature of the information have been provided (References 3, 4, and 5).

PD

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