ML20245B432

Rev 1 to ANF-89-01, Washington Nuclear Power 2 Cycle 5 Plant Transient Analysis

ML20245B432
Person / Time
Site:	Columbia
Issue date:	03/31/1989
From:	Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17285A407	List: ML20245B386 ML20245B432
References
ANF-89-01, ANF-89-01-R01, ANF-89-1, ANF-89-1-R1, TAC-72251, NUDOCS 8904260104
Download: ML20245B432 (52)

Text

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ANF-89-01 '

REVISION 1

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kYj ADVANCED NUCLEAR FUELS CORPORATION a

WNP-2 CYCLE 5 PLANT TRANSIENT ANALYSIS MARCH 1989 g[R04260104 890420 a?OCK 05000397 p

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ADVANCEDNUCLEARFUELS CORPORATION ANF-89-01 i

Revision 1 l

Issue Date:

3/8/89 WNP-2 CYCLE 5 PLANT TRANSIENT ANALYSIS Prepared by

'ta/I.v c Mat c 192 9

() '7 (fJ.E.Krajicek BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services March 1989 I

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NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER I

iMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT 1

PLEASE READ CAREFULLY This technical report was derived through research and development pro-grams sponsored by Advanced Nuclear Fuels Corporation. It is being submit-ted by Advanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Ad-venced Nuclear Fuels Corporation fabricated reload fuel or other tecnnical services provided by Adve'ted Nuclear Fuels Corporation for light water power reactors and it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge,information, and belief. The information con-tanned herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, Dy licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Corporation in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.

Advanced Nuclear Fuels Corporation's warranties and representations con-cerning the subject matter of this document are those set forth in the agree-ment between Advanced Nuclear Fuels Corporation and the customer to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:

A. Makes any warranty, or representation, express or im-plied, with respect to the accuracy, completeness, or usefulness of the information contained in this docu-that the use of any information, apparatus, ment '

method, or process disclosed in this document will not intnnge privately owned rights, or t

B. Assumes any liabilities with respect to the use of, or for i

damages resulting from the use of, any information, ap.

paratus, method. or process disclosed in this document I

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ANF 3las 629A M8i e

4

ANF-89-01 Revision 1 Page i

SUMMARY

OF REVISIONS Revision 1 to ANF-89-01 was issued to address a reload batch size change from 144 to 136 assemblies and minor text changes which describe the reload batch I

size change.,

Feedwater controller failure calculated results at 47% power and 106% flow with normal scram speed and recirculation pump trip are also included for a 144 assembly reload batch size.

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ANF-89-01 Revision 1

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Page ii TABLE OF CONTENTS Section Paae

1.0 INTRODUCTION

I 2.0

SUMMARY

2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN................

5 3.1 Design Basis..........................

5 3.2 Anticipated Transients.....................

5 s

i 3.2.1 Load Rejection Without Bypass..............

6 3.2.2 Feedwater Controller Failure 7

3.2.3 Loss of Feedwater Heating................

8 3.3 Calculational Model 8

3.4 Safety Limit..........................

9 3.5 Final Feedwater Temperature Reduction 9

4.0 MAXIMUM OVERPRESSURIZATION 29 4.1 Design Bases..........................

29 x

4.2 Pressurization Transients 29 4.3 Closure of All Main Steam Isolation Valves...........

29 5.0 RECIRCULATION FLOW RUN-UP......................

31

6.0 REFERENCES

34 APPENDIX A MCPR SAFETY LIMIT A-1 t

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ANF-89-01 Revision 1 Page iii LIST OF TABLES

' Table

' Pace 2.1 THERMAL MARGIN

SUMMARY

FOR WNP-2 CYCLE 5..............

4 3.1 DESIGN REACTOR AND PLANT CONDITIONS FOR WNP-2 10 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 11 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES.............

14 5.1 REDUCED FLOW MCPR OPERATING LIMIT FOR WNP-2 32 LIST OF FIGURES Fiaure Paae

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3.' 1 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED (ORIGINAL RELOAD BATCH SIZE).................

15

/

3.2 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED (0RIGINAL RELOAD BATCH SIZE).................

16 3.3 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED (REVISED RELOAD BATCH SIZE) 17 3.4 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM 18 p

SPEED (REVISED RELOAD BATCH SIZE) 3.5 LOAD REJECTION VITH0VT BYPASS RESULTS, RPT IN0PERABLE, NORMAL 19 SCRAM SPEED............................

3.6 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT IN0PERABLE, NORMAL 20 SCRAM SPEED............................

3.7 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, TECH. SPEC.

21

)

SCRAM SPEED............................

3.8 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, TECH. SPEC.

22 SCRAM SPEED............................

3.9 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT IN0PERARLE, TECH. SPEC.

'3.10 A RJ IONhIiHbViBYPASSRkSbliS,RPiiNbPkRABLE,iECH.'

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SPEC. SCRAM SPEED 24 3.11 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW,

)

RPT OPERABLE, NORMAL SCRAM SPEED..................

25 3.12 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW, RPT OPERABLE, NORMAL SCRAM SPEED..................

26 J

3.13 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW, f

RPT IN0PERABLE, NORMAL SCRAM SPEED.................

27 3.14 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW, L

RPT IN0PERABLE, NORMAL SCRAM SPEED.................

28 L

5.1 REDUCED FLOW MCPR OPERATING LIMIT 33 A.1 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)

A-5 A.2 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL).

A-6 A.3 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL).

A-7 A.4 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 FUEL).

A-8 A.5 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL)....

A9 A.6 RADIAL POWER HIST 0 GRAM FOR 1/4 CORE SAFETY LIMIT MODEL...... A-10

ANF-89-01 Revision 1 Page iv ACKNOWLEDGMENT The author wishes to acknowledge the contribution made to this report by fellow Advanced Nuclear Fuels Corporation employees M. E. Byram, S. J. Haynes, and D. J. Braun.

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ANF-89-01

]

Revision 1 Page 1 1

1.0 INTRODUCTION

This report presents the results of the Advanced Nuclear Fuels Corporation (ANF) evaluation of system transient events for the Supply System Nuclear Project Number 2 (WNP-2) during Cycle 5 operation.

Initially, the analysis the Cycle 5 core was assumed to contain 572 ANF 8x8 and 192 GE P8x8R fuel assemblies.

This document has been revised, at the request of the

~

Washington Public Power Supply System (WPPSS), to reflect a revised Cycle 5 core with eight fewer ANF assemblies or 564 ANF 8x8 and 200 GE P8x8R fuel assemblies.

Since the load rejection without bypass (LRNB) is the limiting pressurization event, only the LRNB event with normal scram speed (NSS) and recirculation pump trip (RPT) operable was recalculated for the revised core loading.

This evaluation together with the analysis of final feedwater temperature reduction (l) (FFTR) and the analysis of core transient events (2) determines the necessary thermal margin (MCPR limits) to protect against boiling transition during the most limiting anticipated operational occurrence (A00).

The evaluation also demonstrates the vessel integrity for the most limiting pressurization event.

This evaluation is applicable for core flows up to the maximum attainable with the recirculation flow control valve in its fully open 70sition which is 106% of the rated core flow value at 100% power.

The methodology used for these system transient analyses is detailed in References 3 and 4.

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ANF-89-01 Revision 1 Page 2 2.0

SUMMARY

The Minimum Critical Power Ratios (MCPR) calculated to protect against boiling transition during potentially limiting plant system transient events are shown in Table 2.1 for powers that bound allowable values.

This table shows the LRNB results for the original and revised reload batch siz.es.

The system transient MCPR values of Table 2.1 for the LRNB and feedwater controller failure (fWCF) transients were obtained using a scram time based on WNP-2 measured values. The loss of feedwater heating (LOFH) transient results

~

s,hown in Table 2.1 were obtained from a bounding analysis which is discussed in Section 3.2.3.

The limiting A00 values for the cases of Table 2.1 are for the LRNB transient at End of Cycle (EOC) conditions; the limiting MCPR values are 1.34 for GE fuel and 1.31 for ANF fuel for the original reload batch size and 1.35 for GE fuel and 1.31 for ANF fuel with the revised reload batch size.

For previous WNP-2 cycles, ANF performed an analysis for the LRNB event at a cycle exposure of E0C -2000 mwd /MTU.

Prior to the end of cycle, a large number of control blades are still inserted in the core.

These. analyses showed that this LRNB system transient was bounded by the control rod withdrawal event (CRWE) by a substantial margiin Thus, for the earlier

)

cycles, plant operating limits were always based on the CRWE for cycle exposures up to E0C -2000 mwd /MTV.

Based on this prior experience, the Cycle 5 MCPR limit up to E0C -2000 has been determined only by the CRWE.(2)

Thus, the Cycle 5 CRWE defined MCPR limit is applicable up to E0C -2000 mwd /MTV, and l

for exposures beyond E0C -2000 mwd /MTU the limits in Table 2.1 are applicable.

l l

Additional transient analyses were performed assuming the recirculation

)r pulp trip (RPT) was out of service, and using the technical specification scram speed (TSSS) and the results are reported herein.

The critical power results for these events are presented in Section 3.0.

The maximum system pressure was calculated for the containment isolation event which is a rapid closure of all main steam isolation valves.

This analysis shows that for WNP-2 Cycle 5 operation, the safety valve response

i ANF-89-01 Revision 1 Page 3 i

j system pressures predicted during the event are below the ASME Code limit of 110% of design pressure (1375 psig) and are shown in Table 2.1.

The analysis conservatively assumed six safety relief valves out of service.

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The continued applicability of the previously established.MCPR safety limit of 1.06 in Cycle 5 was confirmed for all fuel types using the methodology of Reference 6.

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ANF-89-01 Revision 1 Page 4 TABLE 2.1 THERMAL MARGIN

SUMMARY

FOR WNP-2 CYCLE 5 Transient

% Power /% Flow Delta CPR/MCPR*

GE Fuel ANF Fuel Load Rejection **

104/106 0.28/1.34 0.25/1.31 Without Bypass (572 ANF. assemblies &

192 GE assemblies) i Load Rejection **

104/106-0.29/1.35-0.25/1.31 Without Bypass (564 ANF assemblies &

200 GE assemblies Feedwater Controller **

47/106 0.23/1.29 0.20/1.26 Failure Loss of Feedwater***

Not Applicable 0.09/1.15 0.09/1.15 Heating

)

?

MAXIMUM PRESSURE (PSIG)

Transient Vessel Dome Vessel lower Plenum Steam Line MSIV Closure 1286 1315 1289 N:

I

• MCPR value using the 1.06 safety limit justified herein.

• • These transients were evaluated with normal scram speed, RPT operable, and

(-

at the end of cycle.

• • • WNP-2 plant specific bounding value.

).

ANF-89-01 Revision 1 Page 5 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 1

3.1. Desian Basis System analyses were performed at the increased core flow condition of 106% to determine the most limiting type of system transients for the establishment of thermal margins.

As shown in Reference 5, system transients from the increased core flow condition bound transients from the nominal (100%) flow condition. Analysis of the LRNB was performed at the rated design -

104% power /106% flow point.

Since feedwater controller failure (FWCF) transients may be more severe at reduced power because of the larger change in feedwater flow, a FWCF transient was performed at the minimum power (47%) that allowed for increased core flow.

The initial conditions used in the analysis for transients at the 104% power /106% flow point are as shown in Table 3.1.

The most limiting eiposure in cycle was determined to be at end of full power capability when control rods are fully withdrawn from the core; the thermal margin limit established for end of full power conditions is conservative in relation to cases where control rods are partially inserted.

The calculational model s used to analyze these pressurization events include the ANF plant transient and core thermal-hydraulic codes as described in previous documentation.(3,4,5,7)

Fuel pellet-to-clad gap conductances used in the analyses are based on calculations with R0DEX2.(8)

Recirculation pump trip (RPT) coastdown was input based on measured WNP-2 startup test data, and the COTRANSA system transient model for WNP-2 was benchmarked to appropriate i

WNP-2 startup test data.

The hot channel performance is evaluated with XCOBRA-T(4) using COTRANSA supplied boundary conditions. Table 3.2 summarizes the values used for important parameters in the analysis.

l 3.2 Anticipated Transients ANF transient analysis methodology for Jet Pump BWR's considers eight categories of potential system transient occurrences.(3)

The three most l

h limiting transients for WNP-2 are presented in this section; these transients are:

I ANF-89-01 Revision 1

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Page 6 E

Load Rejection Without Bypass (LRNB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LOFH).

A summary of the transient analyses is shown in Table 3.3.

Other plant transient events are inherently nonlimiting or clearly bounded by one of the above events.

3.2.1 Load Re.iection Without Byoass This event is the most limiting of the class of transients characterized by rapid vessel pressurization. The generator load rejection causes a turbine control valve trip, which initiates a reactor scram and a recirculation pump trip (RPT).

The compression wave produced by the fast turbine control valve closure travels through the steam lines into the vessel and pressurizes the s

reactor vessel and core.

Bypass flow to the condenser, which would mitigate the pressurization effect, is conservatively not allowed.

The excursion of core power due to void collapse is primarily terminated by reactor scram and l'

void growth due to RPT.

Figures 3.1 through 3.10 depict the time variance of critical reactor and plant parameters from the analyses of several load rejection transients.

Figures 3.1, 3.2 and 3.5 through 3.10 are load rejection results for the original reload batch size, and Figures 3.3 and 3.4 are load rejection results for the revised reload batch size.

Transient analysis cases include the design basis power and increased cora flow point with a matrix of cases which involve normal scram

speed, technical specification scram speed, and recirculation pump trip (RPT) in service and out of service.

Analysis assumptions are:

Control rod insertion time based on WNP-2 measured data (normal scram speed) or minimum technical specification scram speed.

Integral power to the hot channel was increased by 10% for the pressurization transient, consistent with Reference 9.

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ANF-89-01 Revision 1 Page 7 Table 3.3 shows delta CPR values for a matrix of LRNB transients with the RPT out of service with both normal scram speed (NSS) and technical specification scram speed (TSSS).

ANF has previously analyzed the LRNB event for prior cycles at an exposure of E0C -2000 mwd /MTV. Since a significant number of control rods are inserted into the core up to end-of-cycle (E0C) minus 2000 mwd /MTV, this prior analytical experience has shown the CRWE to be clearly bounding from the beginning-of-cycle (BOC) up to this point.

That is, the limiting delta CPR or MCPR limit throughout the earlier part of the cycle was set by the CRWE from BOC to E0C -2000 mwd /MTU.

For Cycle 5 an LRNB calculation at E0C

-2000 mwd /MTV has not been provided because the CRWE clearly sets the MCPR limit up to this exposure.

For Cycle 5 exposures greater than EOC minus 2000 mwd /HTU, MCPR values deficied in Table 3.3 are applicable.

3.2.2 Feedwater Controller Failure Failure of the feedwater control system is postulated to lead to a maximum increase in feedwater flow into the vessel.

As the excessive l

feedwater flow subcools the recirculating water returning to the reactor core, the core power will rise and attain a new equilibrium if no other action is f

taken.

Eventually, the inventory of water in the downcomer will rise until the high vessel level trip setting is exceeded.

To protect against wet steam l

entering the turbine, the turbine trips upon reaching the high level setting, closing the turbine stop valves. The compression wave that is created, though mitigated by bypass flow, pressurizes the core and causes a power excursion.

The power increase is terminated by reactor scram, RPT, and pressure relief l

from the bypass valves opening.

The evaluation of this event was performed using the scram and integral power assumptions discussed in Section 3.2.1.

Sensitivity results have shown that E0C conditions are bounding because rods

)

are inserted for lower cycle exposures, and high flows are bounding because of higher axials in the core.

Reference 11 showed that the LRNB is more limiting at full power than the FWCF. Because the total change in feedwater flow is the greatest from reduced

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ANF-89-01 Revision 1 Page 8 I

power condition, the FWCF was analyzed from reduced power conditions.

The FWCF was analyzed with the feedwater flow rate increasing at a rate between 10 and 25 percent of nuclear boiler rated (NBR) flow per second. The FWCF transient event was analyzed from the lowest allowed power (47%) at increased core flow.

Figures 3.11 through 3.14 present key variables.

The delta CPR values for the 'co-resident fuel types for 47% power /106% flow transient are E

shown in Table 3.3.

Table 3.3 shows that the delta CPR/MCPR values for the g

FWCF are less than the delta CPR/MCPR value for the 104/106 LRNB event with RPT operable and inoperable with normal scram speed.

3.2.3 Loss of Feedwater Heatina Loss of Feedwater Heating (LOFH) events were evaluated for Cycle 5 with the ANF core simulator model XTGBWR(10) by representing the reactor in l

equilibrium before and after the event.

Actual and projected operating statepoints were used as initial conditions.

Final conditions were determined by reducing the feedwater temperature by 100*F and increasing core power such that the calculated eigenvalue remain unchanged.

I Based on a bounding value analysis, a MCPR limit of 1.15 for WNP-2 with a MCPR safety limit of 1.06 is supported (i.e., a delta CPR of 0.09). As shown in Appendix A of this report, the WNP-2 MCPR safety limit for Cycle 5 continues to be 1.06; hence, the LOFH transient requires a MCPR limit of 1.15 g

for WNP-2.

T l

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3.3 Calculational Model The plant transient codes used to evaluate the pressurization transients (generator load rejection ind feedwater flow increase) were the ANF advanced codes COTRANSA(3) and XCOBRA-T.(4) This axial one-dimensional model predicted reactor power shifts toward the core middle and top as pressurization g

occurred.

This was accounted' for explicitly in determining thermal margin T

changes in the transient.

All pressurization transients were analyzed on a g

bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel 5

model.

The XCOBRA-T code was used consistent with the benchmarking methodology.

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ANF-89-01 Revision 1 Page 9 3.4 Safety limit The MCPR safety limit is the minimum value of the critical power ratio (CPR) at which the fuel could be operated where the expected number of rods in boiling transition would not exceed 0.1% of the fuel rods in the core.

The operating limit MCPR is established such that in the event the most limiting anticipated operational transient occurs, the safety limit will not be violated.

The safety limit for all fuel types in WNP-2 Cycle 5 was confirmed by the s

methodology presented in Reference 6 to have the Cycle 2 value of 1.06.

The input parameters and uncertainties used to establish the safety limit are presented in Appendix A of this report.

3.5 Final Feedwater Temperature Reduction Reference 1

presents final feedwater temperature reduction (FFTR) analysis with thermal coastdown for WNP-2 for Cycles 3 and 4.

The FFTR analysis was performed for a 65'F temperature reduction.

These FFTR analyses are applicable after the all rods out condition is reached with normal feed-water temperature.

The FFTR analysis results show that delta CPR changes for the LRNB and FWCF transients are conservatively bounded by adding 0.02 to the I

delta CPR values for these transients at normal feedwater temperatures.

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I ANF-89-Cl Revision 1 Page 10 Le TABLE 3.1 DESIGN REACTOR AND PLANT CONDITIONS FOR WNP-2 I

Reactor Thermal. Power (104%)

3464 MWt

=

Total Recirculating Flow (106%)

115.0 Mlb/hr Core Channel Flow 107.4 M1b/hr Core Bypass Flow 12.3 Mlb br

/

Core Inlet Enthalpy 527.8 BTU /lbm Vessel Pressures I

Steam Dome 1036. psia Upper Plenum 1049. psia Core 1056. psia Lower Plenum 1073. psia Turbine Pressure 978. psia Feedwater/ Steam Flow 14.8 M1b/hr g

Feedwater Enthalpy 391.1 BTV/lbm Recirculating Pump Flow (per pump) 16.3 M1b/hr i

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E ANF-89-01 Revision 1 Page 11 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 E

High Neutron Flux Trip 126.2%

Void Reactivity Feedback 10% above nominal

• E' Time to Deenergized Pilot Scram 3

Solenoid Valves 200 msec Time to Sense Fast Turbine Control Valve Closure 80 msec Time from High Neutron Flux i

Time to Control Rod Motion 290 msec Normal Tech Spec Scram Insertion Times **

0,404 sec 0.430 sec to Notch 45 0.660 sec 0.868 see to Notch 39 li 1.504 sec 1.936 sec to Notch 25 2.624 sec 3.497 see to Notch 5 Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open l

Turbine Control Valve Stroke Time (Total) 150 msec fuel / Cladding Gap Conductance 2

Core Average (Constant) 587. BTV/hr-ft.p Safety / Relief Valve Performance Settings Technical Specifications Relief Valve Capacity 228.2 lbm/sec (1091 psig) g Pilot Operated Valve Delay / Stroke 400/100 msec 3

• For rapid pressurization transients a 10% multiplier on integral power is used; see Reference 9 for methodology description.

• • Slowest measured average control rod insertion time to specified notches for each group of 4 control rods arranged in a 2x2 array.

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ANF-89-01 Revision 1 Page 12 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)

MSIV Stroke Time 3.0 se MSIV Position Trip Setpoint 85% open Condenser Bypass Valve Performance Total Capacity 990. lbm/sec Delay to Opening (80% open) 300 msec Fraction of Energy Generated in Fuel 0.965 Vessel Water Level (above Separator Skirt)

High Level Trip (L8) 73 in Normal 49.5 in Low Level Trip (L3) 21 in Maximum Feedwater Runout Flow Two Pumps 5799. lbm/sec Recirculating Pump Trip Setpoint 1170 psig Vessel Pressure l

l

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l ANC-89 Revision 1 Page 13 E

TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)

Control Characteristics Sensor Time Constants Steam Flow 1.0 sec g

Pressure 500 msec 5

Others 250 msec Feedwater Control Mode Three-Element Feedwater 100% Mismatch g

Water Level Error 48 in Steam Flow Equiv.

100%

Flow Control Mode Manual Pressure Regulator Settings Lead 3.0 sec Lag 7.0 sec Gain 3.3%/psid E

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ANF-89-01 Revision 1 Page 14 TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES Maximum Maximum Maximum Core Average System Delta CPR Neutron Flux Heat Flux Pressure GE ANF fyv.qni

(% Rated)

(% Rated)

(osia)

Fuel Fuel LRNB 403 121 1169 0.28 0.25 RPT Operable, NSS*

(original reload batch) l LRNB 406 121 1169 0.29 0.25 RPT Operable, NSS (revised reload batch)

LRNB 501 127 1181 0.35 0.31 RPT Inoperable, NSS LRNB 454 127 1174 0.35 0.31 RPT Operable, TSSS**

LRNB 594 132 1189 0.41 0.35 RPT Inoperable, TSSS l

FWCF (47% Power /106%

163 54 1026 0.23 0.20 F'ow),NSS RPT Operable FWCF (47% Power /106%

217 55 1023 0.29 0.25 Flow),NSS RPT Inoperable MSIV Closure With 708 133 1315 N/A Flux Scram NOTES: 1. All results are for the design power and increased flow point (104%

(

power /106% flow) unless otherwise noted.

2. Since there is a small delta CPR increase associated with the LRNB results for the reduced reload batch size, it is conservative to add

)

0.01 to all of the original reload batch size LRNB results to conservatively set limits for the reduced reload batch size.

e

• Normal Scram Speed (NSS).

• • Technical Specification Scram Speed (TSSS).

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ANF-89-01 Revision 1 Page 29 4.0 MAXIMUM OVERPRESSURIZATION Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse

~

scenario as specified by the ASME Pressure Vessel Code.

This analysis showed that the safety valves of WNP-2 have sufficient capacity and performance to prevent pressure from reaching the established transient pressure safety limit of 110% of the design pressure. The maximum system pressures predicted during the event are shown in Table 2.1.

This analysis also assumed six safety relief valves out of service.

=

4.1 Desian Bases

--~

The reactor conditions used in the evaluation of the maximum pressuriza-tion event are those shown in Table 3.1.

The most critical active component (scram on MSIV closure) was assumed to fail during the transient.

The calculation was performed with the ANF advanced plant simulator code 33 COTRANSA,(3) which includes an axial one-dimensional neutronics model.

k 4.2 Pressurization Transients

~

ANF has evaluated several pressurization events and has determined that closure of all main steam itolation valves (MSIVs) without direct scram is the most limiting.

Since the MSIVs are closer to the reactor vessel than the q

turbine stop or turbine control valves, significantly less volume is available 3

to absorb the pressurization phenomena when the MSIVs ere closed than when turbine valves are closed.

The closure rate of the MSIVs is substantially yj s?ower than the turoine stop valves or turbine control valves.

The impact of this smaller volume is more important to this event than the slower closure speed of the MSIV valves relative to turbine valves.

Calculations have determined that the overall result is to cause MSIV closures to be more limiting than turbine isolations.

4.3 Closure of All Main Steam Isolation Valves This calculation also assumed that six relief valves were out of service and that all four main steam isolation valves were isolated at the containment boundary within 3 seconds.

At about 3.3 seconds, the reactor scram is I

1 l

I' l

ANF-89-01 Revision 1 Page 30 I;

initiated by reaching the high flux trip setpoints.

Pressures reach the J

recirculation pump trip setpoint (1170 psig) before the pressurization has been reversed.

Loss of coolant flow leads to enhanced steam production as less subcooled water is available to absorb core thermal power.

The i

calculated maximum pressure in the steam lines was 1289 psig, occurring near the vessel at about 5 seconds.

The maximum vessel pressure was 1315 psig, l

occurring in the lower plenum at about 5 seconds.

These results are presented in Tables 2.1 and 3.3 for the design basis point.

1II Since there has been almost no change in the maximum system pressure i

calculated for the containment isolation event for four cycles, it is reasonable to expect that the reduced reload batch size for Cyle 5 would have l

no impact on the Cycle 5 result given in Tables 2.1 and 3.3.

as )

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I

ANF-89-01 Revision 1 Page 31 5.0 RECIRCULATION FLOW RUN-UP The MCPR full flow operating limit is established through evaluation of anticipated transients at the design basis state.

Due to the potential for large reactor power increases should an uncontrolled recirculation flow increase occur from a less than rated core flow state, the need exists for an augmentation of the operating limit MCPR (full flow) for operation at lower flow conditions.

Advanced Nuclear Fuels Corporation determined the required reduced flow MCPR operating limit by evaluating a bounding slow flow increase event.

The calculations assume the event was initiated from the 104% rod line at minimum flow and terminates at 120% power at 103% flow (flow control valve wide open).

This power flow relationship bounds that calculated for a constant xenon I

assumption.

It was conservatively assumed that the event was quasi-steady and a flow biased scram does not occur.

The power distribution was chosen such that the MCPR equals the safety limit at the final power / flow run-up point.

The reduced flow MCPRs were then calculated by XCOBRA(6) at discrete flow points.

I The recirculation flow run-up analysis performed for WNP-2 Cycle 2 was reviewed, and the assumptions and conditions used for Cycle 2 are applicable to Cycle 5 except for the six degree reduction in feedwater temperature at full power conditions.

Thus, the reduced flow MCPR operating limit for WNP-2 Cycle 5 is changed slightly from earlier cycl es.

For final feedwater temperature reduction (FFTR) conditions, the previously reported (l) reduced flow MCPR operating limit remai n applicable. The reduced flow MCPR operating limit for Cycle 5 is presented in Figure 5.1 and tabulated in Table 5.1.

The MCPR operating limit for WNP-2 shall be the maximum of this reduced flow MCPR I

operating limit and the full flow MCPR operating limit as summarized in Reference 2.

I I

ANF-89-01 Revision 1 Page 32 TABLE 5.1 REDUCED FLOW MCPR OPERATING LIMIT FOR WNP-2 Core Flow Reduced Flow MCPR

(% Rated.1

__Qoeratina Limit 100 1.07 90 1.13 80 1.19 70 1.26 60 1.34 50 1.44 40 1.59 I

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ANF-89-01 Revision 1 Page 34 N

6.0 REFERENCES

I 1.

J.

E.

Krajicek, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction,"

XN-NF-87-92 and XN-NF-87-92, Supplement 1,

Advanced Nuclear fuels Corporation, Richland, WA 99352, June 1987 and May 1988.

2.

J. E. Krajicek, " Supply System Nuclear Project Number 2 (WNP-2) Cycle 5 Reload Analysis,"

ANF-89-02, Revision 1,

Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1989.

3.

R.

H.

Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(P), Revision 2 (as supplemented), Exxon I

Nuclear Company, Inc., Richland, WA 99352, November 1981.

4.

M. J. Ades and B. C. Fryer, "XCOBRA-T: A Computer Code for BWR Transient I

Thermal-Hydraulic Core Analysis," XN-NF-84-105(A), Volume 1, Volume 1 Supplement 1,

Volume 1 Supplement 2 and XN-NF-84-105(A), Volume 1 Supplement 4, Advanced Nuclear Fuels Corporation, Richland, WA 99352, February 1987 and July 1987.

5.

J.

B.

Edgar, Letter to WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear

Company, Inc.,

Richland, WA

99352, i

April 15, 1986.

a 6.

T. W. Patten, " Exxon Nuclear Critical Power Methodology for Boiling Water g

Reactors,"

XN-NF-524(A),

Revision 1,

Exxon Nuclear Comoany, Inc.,

Richland, WA 99352, November 1983.

7.

T.

L.

Krysinski and J.

C.

Chandler, " Exxon Nuclear Methodology for Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary Description,"

XN-NF-80-19(A),

Volume 3,

Revision 2,

Exxon Nuclear Company, Inc., Richland, WA 99352, January 1987.

8.

K.

R.

Merckx, "RODEX2 Fuel Rod Mechanical Response Evaluation Model,"

XN-NF-81-58(A),

Revision 2,

Exxon Nuclear

Company, Inc.,

Richland, WA 99352, March 1984.

9.

S.

E.

Jensen, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors:

Revised Methodology for Including Code Uncertainties in I

Determining Operating Limits for Rapid Pressurization Transients in BWRs," XN-NF-79-71( A), Revision 2, Supplements 1, 2, and 3, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1986.

10.

" Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods for Design Analysis," XN-NF-80-19(A), Volume 1,

Supplements 1 and 2, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.

I I

I ANF-89-01 Revision 1 Page 35 I

11.

J. E. Krajicek, "WNP-2 Cycle 2 Plant Transient Analysis," XN-NF-85-143, Revision 1,

Exxon Nuclear

Company, Inc.,

Richland, WA

99352, February 1986.

I I

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ANF-89-01 Revision 1 Page A-1 APPENDIX A MCPR SAFETY LIMIT A.1 INTRODUCTION Bundle power limits in a boiling' water reactor (BWR) are determined through evaluation of critical heat flux phenomena.

The basic criterion used in establishing critical power ratio (CPR) limits is that at least 99.9% of the fuel rods in the core will be expected to avoid boiling transition (critical heat flux) during normal operation and anticipated operational occurrences.

Operating margins are defined by establishing a minimum margin to the onset of boiling transition condition for steady state operation and calculating a transient effects allowance, thereby assuring that the steady state limit is protected during anticipated off-normal conditions.

This appendix addresses the calculation of the minimum margin to the steady state boiling transition condition, which is implemented as the MCPR safety limit in the plant technical specifications.

The transient effects allowance, or the limiting transient change in CPR (i.e., delta CPR), is treated in the body of this report.

The MCPR safety limit is established through statistical consideration of measurement and calculational uncertainties associated with the thermal hydraulic state of the reactor using design basis radial, axial, and local power distributions. Some of the calculational uncertainties, including those introduced by the critical power correlation, power peaking, and cere coolant l

distribution, are fuel related. When ANF fuel is introduced into a core where it will reside with another supplier's fuel types, the app.priate value of the MCPR safety limit is calculated based on fuel-dependent parameters associated with the mixed core.

Similarly, when an ANF-fabricated reload l

batch is used to replace a group of dissimilar fuel assemblies, the core I

average fuel dependent parameters change because of the difference in the relative number of each type of bundle in the core, and the MCPR safety limit is again reevaluated.

i

I ANF-89-01 Revision 1 1

Page A-2

{

The design basis power distribution is made up of components corresponding to representative radial, axial, and local peaking factors.

Where such data are appropriately available from the previous cycle, these factors are determined through examination of operating data for the previous cycle and predictions of operating conditions during the cycle being evaluated for the MCPR safety limit.

If operating data are not available, either g

because the reactor has not been operated or because appropriate data cannot 3

be supplied to ANF, the safety limit power distribution is determined strictly from the predicted operating conditions during the cycle being evaluated.

Operating data for WNP-2 during Cycle 4 and the predicted operating conditions for Cycle 5 were evaluated to identify the design basis power distributions used in the Cycle 5 MCPR safety limit analysis.

I E

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ANF-89-01 Revision 1 Page A-3 9

.A.2 ASSUMPTIONS A.2.1 Desian Basis Power Distribution The local and radial power distributions which were determined to be conservative. for use in the ' safety limit analysis are shown in Figures A-1 through A-5.

A.2.2 Hydraulic Demand Curve Hydraulic demand curves' based on calculations with.XCOBRA.were used in the safety limit analysis. The XCOBRA calculation is described in ANF topical reports XN-NF-79-59(A), " Methodology for Calculation of Pressure Drop in BWR Fuel-Assemblies," and XN-NF-512(A), "The XN-3 Critical Power Correlation."

A. 2.3 ~ _S; stem Uncertainties System measurement uncertainties are not fuel dependent.

The values

-reported-by the NSSS supplier for these parameters remain valid for the insertion of ANF fuel.

The values used in the ' safety limit analysis are tabulated in the topical report XN-NF-524(A), " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."

A.2.4 Fuel Related Uncertainty E fL Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty.

The vdues used in the safety limit analysis are tabulated in the topical report XN-NF-524(A), " Exxon Nuclear Critical i

Power Methodology for Boiling Water Reactors."

Power measurement uncertainties are established in the topical report XN-NF-80-19(A), Volume 1,

" Exxon Nuclear Methodology for Boiling Water Reactors; Neutronics Methods for Design and Analysis."

)

s ANF-89-01

.i Revision 1 Page A-4 A.3 SAFETY LIMIT CALCULATION A statistical analysis.for the number of fuel rods in boiling transition was performed using the methodology described in ANF topical report XN-NF-524(A), " Exxon Nuclear Critical Power Methodology for Boiling Water -

Reactors."

With 500 Monte Carlo trials it was determined that for a minimum CPR value of 1.06 at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95%.

l

ANF-89-01 Revision 1

.Page A-5

.936 :

.977 : 1.023 : 1.015 : 1.011 : 1.041 : 1.076 : 1.052 :

.977 : 1.011 :

.907 : 1.042 : 1.035 :

.932 :

.962 : 1.075':

1.023 :

.907 : 1.017 :

.988 :

.974 :

.996 :

.931 : 1.040 :

1.015 : 1.042.:

.988 :

.000 :

.850 :

.972 : 1.033 : 1.009 :

1.011 : 1.035 :

.974 :

.850 :

.000 :

.985 : 1.038 : 1.011 :

1.041 :

.932 :

.996 :

.972 :

.985 : 1.012 :

.901 : 1.043 :

l-

1.076 :

.962 :

.931.: 1.033 : 1.038 :

.901 :

.976 : 1.078 :

1.052 : 1.075 : 1.040 : 1.009 : 1.011 : 1.043 : 1.078 : 1.054 :

r FIGURE A.1 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)

I ANF-89-01 Revision 1 Page A-6 I

i

.944 :

.962 : 1.011 : 1.044 : 1.043 : 1.010 :

.960 :

.943 :

.962 :

.980 : 1.064 :

.894 : 1.033 : 1.059 : 1.034 :

.961 :

1.011 : 1.064 : 1.010 :

.994 :

.982 : 1.002 :

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.000 :

.907 :

.980 : 1.032 : 1.042 :

J j

1

1.045 : 1.033 :

.982 :

.907 :

.000 :

.988 :

.952 : 1.041 :

l

)

1.010 : 1.059 : 1.002 :

.980 :

.988 : 1.004 : 1.060 : 1.065 :

.960 : 1.034 :

.915 : 1.032 :

.952 : 1.060 :

.966 : 1.053 :

l j

.943 :

.961 : 1.010 : 1.042 : 1.041 : 1.065 : 1.053 : 1.019 :

l 1

I!

FIGURE A.2 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS I

l (ANF XN-3 FUEL) l ll I

L

ANF-89-01 Revision 1 Page A-7

.950 :

.963 : 1.000 : 1.027 : 1.026 :

.999 :

.963 :

.950 :

.963 :

.981 : 1.052 :

.920 : 1.033 : 1.049 : 1.020 :

.963 :

1.000 : 1.052 : 1.017 : 1.005 :

.997 : 1.011 :

.936 : 1.000 :

1.027 :

.920 : 1.005 :

.000 :

.935 :

.996 : 1.033 : 1.027 :

• : 1.026 : 1.033 :

.997 :

.935 :

.000 : 1.002 :

.971 : 1.027 :

.999 : 1.049 : '1.011 :

.996 : 1.002 : 1.016 : 1.054 : 1.042 :

l l

.963 : 1.020 :

.936 : 1.033 :

.971 : 1.054 :

.973 : 1.029 :

l

.950 :

.963 : 1.000 : 1.027 : 1.027 : 1.042 : 1.029 : 1.003 :

1 e

l FIGURE A.3 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)

I ANF-89 01 Revision 1 Page A-8 I

i

.967 :

.969 :

.997 : 1.019 : 1.019 :

996 :

.968 :

.966 :

I

.969 :

.981 : 1.044 :

.932 : 1.030 : 1.042 : 1.013 :

.968 :

.997 : 1.044 : 1.017 : 1.008 : 1.001 : 1.012 :

.944 :

.997 :

1.019 :

.932 : 1.008 :

.000 :

.947 : 1.000 : 1.030 : 1.019 :

k

.e e

1.019 : 1.030 : 1.001 :

.947 :

.000 : 1.006 :

.976 : 1.020 :

.996 : 1.042 : 1.012 : 1.000 : 1.006 : 1.017 : 1.047 : 1.032 :

.968 : 1.013 :

.944 : 1.030 :

.976 : 1.047 :

.975 : 1.020 :

.966 :

.968 :

.997 : 1.019 : 1.020 : 1.032 : 1.020 : 1.003 :

.......................e.P.............em..

as.........e

....es..m....

I l

FIGURE A.4 WNP '2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS i

(ANF XN-1 FUEL)

I I

i

. 1 ANF-89-01 Revision 1 Page A-9

.1 1

4 l

1.03':

1.00 :

.99 :

.99 :

.99 :

.99 :

1.00 :

1.03 :

1.00 :

.97 :

.99 :

1.02 :

1.03 :

1.03 :

.99 :

1.00 :

i

.99 :

.99 :

1.02 :

1.01 :

1.02 :

.91 :

1.03 :

.99 :

.99 :

1.02 :

1.01 :

.91 :

.00 :

1.02 :

1.02 :

.99 :

.99 :

1.03 :

1.02 :

.00 :

1.02 :

1.01 :

.99 :

.99 :

.99 :

1.03 :

.91 :

1.02 :

1.01 :

.98 :

.99 :

.99 :

)

1.00 :

.99 :

1.03 :

1.02 :

.99 :

.99 :

.97 :

1.00 :

l 1.03 :

1.00 :

.99 :

.99 :

.99 :

.99 :

1.00 :

1.03 :

t l

FIGURE A.5 WNP-2 CYCLE 5 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL)

l I1l s

5 s-2[

i s

7 1

LE D

n O

0 M

s T

I 1

M I

L YTE sR F

O A

2. T S

1 C

E l

R f

O F

C R

/

4 E

1 iW R

O O

P F

M E

AR L

G sD O

7. N T

S 0U I

B H

R EW O

0 P

s.

L

~

0 A

I DA R

s 6

2 A

V 0

E R

U n

G I

F 0

0 0

0 0

0 0

8 6

4 2

1

!aODm wO xyrDZ 3

l 1'

t 4l c

dNF-89-01 ~

Revision 1 Issue Date: 3/8/89 i

WNP-2 CYCLE 5 PLANT TRANSIENT ANALYSIS i

Distribution:

0. C. Brown M. E. Byram-R. E. Collingham S.-J. Haynes' S. E. Jensen J. E. Krajicek S. L. Leonard J. L. Maryott L. A. Nielsen G. L. Ritter R.- B. Stout /D. C. Kilian H. E. Williamson Y. U. Fresk/WPPSS (51).

Document Control (5) i

.