ML17285B057

Washington Nuclear Plant-2 Cycle 6 Plant Transient Analysis.

ML17285B057
Person / Time
Site:	Columbia
Issue date:	01/31/1990
From:	Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17285B053	List: ML17285B052 ML17285B054 ML17285B055 ML17285B056 ML17285B057
References
ANF-90-01, ANF-90-1, NUDOCS 9003080240
Download: ML17285B057 (58)

Text

ADVANCEDNUCLEARFUELS CORPORATION ANF-90-01 Issue Date: 1/10/90 WNP-2 CYCLE 6 PLANT TRANSIENT ANALYSIS Prepared by J. E. Krajicek WR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services January 1990 90 3080 DR qgOC 9002p~-

050003yy POC

NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT 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 Commlssjon as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Ad-vanced Nuclear Fuels Corporation.fabricated reload fuel or other technical services provided by Advanced 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.

talned 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, by 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.

ment, or that the use of any Information, apparatus, method, or process disclosed In this document will not Infringe privately owned rights, or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any Information, ap.

paratus, method. or process disclosed In this document.

ANF4 mls 629A &88)

ANF-90-01 Page i TABLE OF CONTENTS Section Pacae

1. 0 INTRODUCTION 1 2.0

SUMMARY

. 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN . 5

3. 1 Design Basis . 5 3.2 Anticipated Transients . 5 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 Hodel 8 3.4 Safety Limit . . . . . . . . . . . . . 9 3.5 Final Feedwater Temperature Reduction 9

4. 0 MAXIMUM OVERPRESSUR IZATION 27 4.1 Design B'ases . 27 4.2 Pressurization Transients 27 4.3 Closure of All Hain Steam Isolation Val ves . 27 5.0 RECIRCULATION FLOW RUN-UP . ~ ~ ~ 29

6.0 REFERENCES

30 APPENDIX A HCPR SAFETY LIMIT A-1

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ANF-90-01 Page ii LIST OF TABLES Table Pacae 2.1 3.1 3.2 THERMAL MARGIN

SUMMARY

FOR WNP-2 CYCLE 6 RELOAD REACTOR AND PLANT ANALYSIS CONDITIONS FOR WNP-2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 FUELS........

...... 10 11 4

3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES............. 14 LIST OF FIGURES

~Fi use ~Pa e 3.1 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM 15 3.2 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM 16 3.3 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, NORMAL SCRAM SPEED ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ t ~ ~ ~ ~ ~ ~ 17 3.4 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, NORMAL SCRAM SPEED 3.5 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, TECH. SPEC. 18'9'0 SCRAM SPEED 3.6 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, TECH. SPEC.

SCRAM SPEED 3.7 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, TECH. SPEC.

SCRAM SPEED 21 3.8 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT INOPERABLE, TECH. SPEC.

SCRAM SPEED 22 3.9 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW RPT OPERABLE, NORMAL SCRAM SPEED . 23 3 10 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW RPT OPERABLE, NORMAL SCRAM SPEED . 24 3 ll FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW RPT INOPERABLE, NORMAL SCRAM SPEED . 25

3. 12 FEEDWATER CONTROLLER FAILURE RESULTS FOR 47% POWER AND 106% FLOW RPT INOPERABLE, NORMAL SCRAM SPEED . 26 A. 1 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-5 FUEL) A-5 A.2 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF 4 FUEL) ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A-6 A.3 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL) . . . . . . . . . . . . . . . . . . . . ~ A-7 A.4 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL) A-8

ANF-90-01 Page iii A.5 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 CENTRAL FUEL) A-9 A.6 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 PERIPHERAL FUEL) A-10 A.7 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL) A-ll A.S RADIAL POWER HISTOGRAM FOR I/O CORE SAFETY LIMIT MODEL A-12

ANF-90-01 Page iv ACKNOWLEDGMENT The author wishes to acknowledge the contribution made to this report by fellow Advanced Nuclear Fuels Corporation employees: L. G. Riniker, M. L. Hymas, D. J. Braun, D. F. Richey and S. L. Leonard.

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ANF-90-01 Page v WNP-2 CYCLE 6 PLANT TRANSIENT ANALYSIS REPORT The following Design Notebooks support the above titled document.

E-5003-593-1 WNP-2 Cycle 6 XCOBRA Data DECKPL E-5003-593-2 WNP-2 Cycle 6 Heat Transfer Coefficients E-5003-593-3 WNP-2 Cycle 6 XCOBRA-T Data DECKPL E-5003-593-4 WNP-2 Cycle 6 COTRANSA Input Deck Preparation E-5003-595-1 WNP-2 Cycle 6 104/106 LRNB RPT Operable NSS E-5003-595-2 WNP-2 Cycle 6 104/106 LRNB RPT Inoperable NSS E-5003-595-3 WNP-2 Cycle 6 104/106 LRNB RPT Operable TSSS E-5003-595-4 WNP-2 Cycle 6 104/106 LRNB RPT Inoperable TSSS E-5003-596-5 WNP-2 Cycle 6 47/106 FWCF RPT Operable NSS E-5003-596-6 WNP-2 Cycle 6 47/106 FWCF RPT Inoperable NSS E-5003-595-7 WNP-2 Cycle 6 ASME Overpressurization Transient E-5003-872-1 WNP-2 Cycle 6 MCPR Safety Limit Analysis

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ANF-90-01 Page 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 6 operation. This analysis is for the Cycle 6 core was assumed to contain 708 ANF 8x8 and 56 GE P8x8R fuel assemblies.

This evaluation is applicable for core flows up to the maximum attainable with the recirculation flow control valve in its fully open position which is

- 106% of the rated core flow value at 100% power. The methodology used for these system transient analyses is detailed in References 1 and 2. The results are used in a subsequent Reload Analysis Report( ) to justify plant operating limits and set points.

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ANF-90-01 Page 2 2.0

SUMMARY

The Minimum Critical Power Ratios (HCPR) calculated to protect against boiling transition during potentially limiting plant system transient events are shown in Table 2. 1. The system transient HCPR values of Table 2. 1 for the load rejection without bypass (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 shown in Table 2. 1 were obtained from a bounding analysis which is discussed in Section 3.2.3. The limiting AOO event MCPR values for the cases of Table 2. 1 are for the LRNB transient at End of Cycle (EOC) conditions. The limiting MCPR value is 1.31 for ANF fuel, and this HCPR limit also bounds the GE fuel in the core periphery. Results for GE fuel in Cycle 6 are shown in Table 3.3 of this report.

Analyses for earlier cycles showed that the LRNB system transient at a cycle exposure of EOC - 2000 HWd/HTU was bounded by the control rod withdrawal event (CRWE) by a substantial margin. Based on this prior experience, the MCPR limit for Cycle 6 up to EOC -2000 have been determined by the CRWE.(3)

Additional transient analyses were performed assuming the recirculation pump trip (RPT) was out of service and using the technical specification scram speed (TSSS). The results from these analyses are also 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 6 operation, the safety valve response prevents the pressure from reaching 110% of design pressure. The maximum 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.

ANF-90-01 Page 3 The continued applicability of the previously established HCPR safety limit of 1.06 in Cycle 6 was confirmed for all fuel types in the Cycle 6 loading using the methodology of Reference 6.

ANF-90-01 Page 4 TABLE 2.1 THERMAL MARGIN

SUMMARY

FOR WNP-2 CYCLE 6 RELOAD FUELS Transient K~F1 Delta CPR HCPR*

Load Rejection** 104/106 0.25/1.31 Without Bypass Feedwater Controller** 47/106 0.21/1.27 Failure Loss of Feedwater*** Not Applicable 0.09/1.15 Heating MAXIMUM PRESSURE (PSIG)

Transient Vessel Dome Vessel Lower Plenum Steam Line HSIV Closure 1289 1317 1291

• 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.

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ANF-90-01 Page 5 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 D~iB 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 4, 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 exposure 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 models used to analyze these pressurization events include the ANF plant transient and core thermal-hydraulic codes as described in previous 'documentation.( , >> ) Fuel pellet-to-clad gap conductances used in the analyses are based on calculations with RODEX2 for exposures at which the fuel is calculated to be limiting.( ) 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 WNP-2 startup test data. The hot channel performance is evaluated with XCOBRA-T( )

using COTRANSA supplied boundary conditions. Table 3.2 summarizes the values used for important parameters in the analysis.

3.2 Antici ated Transients ANF transient analysis methodology for Jet Pump BWRs considers eight categories of potential system transient occurrences.( ) The three most limiting transients for WNP-2 are presented in this section; these transients

ANF-90-01 Page 6 Load Rejection Without Bypass (LRNB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LOFH).

A summary of the transient analyses results is shown in Table 3.3. The delta CPR values for ANF Sx8 fuel bound the co-resident GE Sx8 fuel for the LRNB and FWCF transients. Other plant transient events are inherently nonlimiting or clearly bounded by one of the above events.

3.2. 1 Load Re 'ection Without B ass 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 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 void growth due to RPT. Figures 3. 1 through 3.8 depict the time variance of critical reactor and plant parameters from the analyses of several load rejection transients. Transient analysis cases include the design basis power and increased core 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 8.

ANF-90-01 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 analyzed the LRNB event for Cycles 2 through 4 at an exposure of EOC -2000 MWd/HTU. Since a significant number of control rods are inserted into the core up to end-of-cycle (EOC) minus 2000 MWd/MTU, this prior analytical experience has shown the CRWE'o be clearly bounding from the beginning-of-cycle (BOC) up to this point. That is, the limiting delta CPR or HCPR limit throughout the earlier part of the cycle was set by the CRWE from BOC to EOC -2000 MWd/HTU. For Cycles 5 and 6 an LRNB calculation at EOC

-2000 HWd/HTU has not been provided because the CRWE clearly sets the HCPR limit up to this exposure. For Cycle 6 exposures greater than EOC minus 2000 HWd/MTU, MCPR values defined 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 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 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 entering the turbine, the turbine trips upon reaching the high level setting, closing the turbine stop valves. Credit is taken for the control system signal to open the bypass 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 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 EOC conditions are bounding because rods are inserted for lower cycle exposures, and high flows are bounding because of axial power peaking higher in the core.

ANF-90-01 Page 8 Reference ll 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 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 3.9 through 3. 12 present key variables. The results are shown in Table 3.3. Table 3.3 shows that the delta CPR/HCPR values for the FWCF are less than the delta CPR/HCPR value for the 104/106 LRNB event with RPT operable and inoperable for normal scram speed.

3.2.3 Loss of Feedwater Heatin Loss of Feedwater Heating (LOFH) events were evaluated for Cycle 6 with the ANF core simulator model XTGBWR( ) by representing the reactor in 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 remains unchanged.

Based on a bounding value analysis, a HCPR 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 HCPR safety limit for Cycle 6 continues to be 1.06; hence, the LOFH transient requires a HCPR limit of 1. 15 for WNP-2.

3.3 Calculational Nodel The plant transient codes used to evaluate the pressurization transients (generator load rejection and feedwater flow increase) were the ANF advanced codes COTRANSA( ) and XCOBRA-T.(2) This axial one-dimensional model predicted reactor power shifts toward the core middle and top as pressurization occurred. This was accounted for explicitly in determining thermal margin changes in the transient. All pressurization transients were analyzed on a bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel

ANF-90-01 Page 9 model. The XCOBRA-T code was used consistent with the benchmarking methodology.

.4 5 F Li The HCPR 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 F boiling transition would not exceed 0. 1% of the fuel rods in the core. The operating limit HCPR 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 6 was confirmed by the methodology presented in Reference 5 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 Tem erature Reduction Reference 10 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 delta CPR values for these transients at normal feedwater temperatures.

ANF-90-01 Page 10 TABLE 3.1 REACTOR AND PLANT ANALYSIS CONDITIONS FOR WNP-2 Reactor Thermal Power (104%) 3464 MWt Total Recirculating Flow (106%) 115.0 Hlb/hr Gore Channel Flow 103.0 Mlb/hr Core Bypass Flow 12.0 Hlb/hr Core Inlet Enthalpy 527.8 BTU/ibm Vessel Pressures Steam Dome 1036. psia Upper Plenum 1049. psia Core 1056. psia Lower Plenum 1072. psia Turbine Pressure 979. psia Feedwater/Steam Flow 14.6 Hlb/hr Feedwater Enthalpy 391.1 BTU/1 bm Recirculating Pump Flow (per pump) 17.3 Mlb/hr

ANF-90-01 Page 11 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 High Neutron Flux Trip 126.2%

Void Reactivity Feedback 10% above nominal*

Time to Deenergized Pilot Scram Solenoid Valves 200 msec Time to Sense Fast Turbine Control Valve Closure 80 msec'80 Time from High Neutron Flux Time to Control Rod Motion msec Normal Tech Spec Scram Insertion Times** 0.404 sec 0.430 sec to Notch 45 0.660 sec 0.868 sec to Notch 39 1.504 sec 1.936 sec to Notch 25 2.624 sec 3.497 sec to Notch 5 Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open Turbine Control Valve Stroke Time (Total) 150 msec Fuel/Cladding Gap Conductance Core Average (Constant) 584. BTU/hr-ft2-F Safety/Relief Valve Performance Settings Technical Specifications Relief Valve Capacity 226.0 ibm/sec (1106 psig)

Pilot Operated Valve Delay/Stroke 400/100 msec

• 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.

ANF-90-01 Page 12 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)

MSIV Stroke Time 3.0 sec MSIV Position Trip Setpoint 85% open Condenser Bypass Valve Performance Total Capacity 990. ibm/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 ln Normal 49.5 in Low Level Trip (L3) 21 in Maximum Feedwater Runout Flow Two Pumps 5686. ibm/sec Recirculating Pump Trip Setpoint 1170 psig Vessel Pressure

ANF-90-01 Page 13 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)

Control Characteristics Sensor Time Constants Steam Flow 1.0 sec Pressure 500 msec Others 250 msec Feedwater Control Mode Three-Element Feedwater 100% Mismatch 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

ANF-90-01 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 Event  % Rated  % Rated ~sic/ Fuel Fuel LRNB 397 117 1203 0.24 0.25 RPT Operable, NSS*

LRNB 553 122 1212 0.28 0.30 RPT Inoperable, NSS LRNB 483 122 1205 0.27 0.30 RPT Operable, TSSS**

LRNB 592 132 1189 0.31 0.34 RPT Inoperable, TSSS FWCF (47% Power/106% 173 55 1026 0.20 0.21 Flow), NSS RPT,Operable FWCF (47% Power/106% 230 57 1037 0.23 0.25 Flow), NSS RPT Inoperable HSIV Closure With 813 136 1317 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. ANF 8x8 fuel results bound GE 8x8 fuel results in all cases; hence, only ANF fuel results need to be reported in the Technical Specifications for Cycle 6.

• Normal Scram Speed (NSS).

• • Technical Specification Scram Speed (TSSS).

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ANF-90-01 Page 27

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 ASHE 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 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 COTRANSA,( ) which includes an axial one-dimensional neutronics model.

4.2 Pressurization Transients ANF has evaluated several pressurization events and has determined that closure of all main steam isolation valves (HSIVs) without direct scram is the most limiting. Since the HSIVs are closer to the reactor vessel than the-turbine stop or turbine control valves, significantly less volume is available to absorb the pressurization phenomena when the HSIVs are closed than when turbine valves are closed. The closure rate of the MSIVs is substantially slower than the turbine stop valves or turbine control valves. The impact of this smaller volume is more important to this event than the slower closure spied of the HSIV valves relative to turbine valves. Calculations have determined that the overall result is to cause HSIV closures to be more limiting than turbine isolations.

4.3 Closure of All Hain 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

ANF-90-01 Page 28 initiated by reaching the high flux trip setpoints. Pressures reach the 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 calculated maximum pressure in the steam lines was 1289 psig, occurring near the vessel at about 5 seconds. The maximum vessel pressure was 1317 psig, 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.

ANF-90-01 Page 29 5.0 RECIRCULATION FLOW RUN-UP The HCPR 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 HCPR (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 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 HCPR equals the safety limit at the final power/flow run-up point. The reduced flow HCPRs were then calculated by XCOBRA(5) at discrete flow points.

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 Cycles 5 and 6 except for the six degree reduction in feedwater temperature at full power conditions. Thus, the reduced flow HCPR operating limit for WNP-2 Cycle 5 was changed slightly from earlier cycles. For final feedwater temperature reduction (FFTR) conditions, the previously reported(I ) reduced flow HCPR operating limit remains applicable. The FFTR reduced flow HCPR operating limit calculated for Cycle 5( ) remains applicable to Cycle 6. The HCPR operating limit for WNP-2 shall be the maximum of this Cycle 5 calculated reduced flow HCPR operating limit and the full flow MCPR operating limit as summarized in Reference 3.

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ANF-90-01 Page 30

6.0 REFERENCES

R.

N H.

R," ~X-NF->>-I, Nuclear Company, Inc., Richland, WA R 11 2 (>>I Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling 99352, November 1981.

d), E

2. H.

it J.

Eppl Ades and B. C.

I-Ryd Supplement 4, 11 I,VI 2 Advanced Fryer, A

I dppi "XCOBRA-T:

ly I,"

Nuclear Fuels A

~XN-NF-t

-I, Computer Code 2d~X-F--,V11 Corporation, V I for Richland, BWR I,

Transient WA V I 99352, February 1987 and July 1987.

3. J. E. Krajicek, "WNP-2 Cycle 6 Reload Analysis," ANF-90-02, Advanced Nuclear Fuels Corporation, Richland, WA 99352, January 1990.

J. B. Edgar, Letter to WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.

5. T.

R t,"

W. Patten, Richland, WA "Exxon Nuclear

~XN-Np- VA, R 99352, November 1983.

Critical 11 Power Methodology I, E N I for Boiling C p Water "Exxon Nuclear Methodology for

6. T.

P L.

ip Krysinski I,"

Boiling Water Reactors; X~F-and Inc., Richland, J. C. Chandler, THERHEX V I Thermal Limits Methodology; Summary 99352, January 1987.

I, R 11 2, E N I-,

Company, WA

7. K. R. Herckx, "RODEX2 Fuel Rod Mechanical Response Evaluation Model,"

~XN-IIF- R 11 I., E II I C p Richland, WA 99352, March 1984.

8. S. E. Jensen, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors: Revised Hethodology for Including Code Uncertainties in Determining Operating Limits for Rapid Pressurization Transients in XNR,"~K-F-7-7 A,R 11 I.,Eppl 1,2, ER,E II Company, Inc., Richland, WA 99352, March 1986.

9. "Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods I p Id A I I," ~XN-NF-8-I9A, V I I, 2 pl t I d 2, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.

10. 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 Hay 1988.

J. E. Kr ajicek, "WNP-2 Cycle 2 Plant Transient Analysis," XN-NF-85-143, Revision 1, Exxon Nuclear Company, Inc , Richland, WA 99352, February 1986.

ANF-90-01 Page 31

12. J. E. Krajicek, "WNP-2 Cycle 5 Plant Transient Analysis," ANF-89-01, Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, Mar ch 1989.

ANF-90-01 Page A-1 APPENDIX A HCPR 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 HCPR 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 core coolant distribution, are fuel related. When ANF fuel is introduced into a core where it will reside with another supplier's fuel types, the appropriate value of the HCPR safety limit is calculated based on fuel-dependent parameters associated with the mixed core. Similarly, when an ANF-fabricated reload batch is used to replace a group of dissimilar fuel assemblies, the core average fuel dependent parameters change because of the difference in the relative number of each type of bundle in the core, and the HCPR safety limit is again reevaluated.

ANF-90-01 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. Operating data for WNP-2 during Cycle 5 and the predicted operating conditions for Cycle 6 were evaluated to identify the design basis power distributions used in the Cycle 6 HCPR safety limit analysis.

0 ANF-90-01 Page A-3 A. 2 ASSUMPTIONS A.2. 1 Desi n 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 H draulic 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 I

tabulated in the topical report XN-NF-524(A), "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."

A.2.4 Fuel Related Uncertainties Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty. 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." 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."

ANF-90-01 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%.

ANF-90-01 Page A-5

• ~

• : 0.936  : 0.977 ': 1.023  : 1.015 : 1.011 : 1.041  : 1.076 : 1.052 :

• ~

0.977  : 1.011  : 0.907  : 1.042 : 1.035  : 0.932 : 0.962 : 1.075 :

• ~

• ~

• : 1.023  : 0.907  : 1.017  : 0.988 : 0.974  : 0.996 : 0.931 : 1.040 :

• ~

• ~

• : 1.015  : 1.042  : 0.988  : 0.000 : 0.850  : 0.972 : 1.033 : 1.009 :

• ~

• : 1.011  : 1.035  : 0.974  : 0.850 : 0.000  : 0.985 : 1.038 : 1.011

• ~

• : 1.041  : 0.932 : 0.996  : 0.972 : 0.985  : 1.012 : 0.901 : 1.043 :

• ~

1.076  : 0.962 : 0.931  : 1.033 : 1.038  : 0.901 : 0.976 : 1.078 :

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

FIGURE A.1 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-5 FUEL)

I ~ ~

ANF-90-01 Page A-6

• ~

• : 0.928  : 0.958  : 1.002 : 1.004 : 1.002  : 1.019 : 1.032 : 1.000 :

• ~

0

• ~

• : 0.958  : 0.998  : 0.938  : 1.047 : 1.042  : 0.952 : 0.974 : 1.032 :

• ~

• : 1.002  : 0.938.; 1.032  : 1.015  : 1.002  : 1.015 : 0.952 : 1.019 :

• ~

• ~

• : 1.004 : 1.047  : 1.015  : 0.000 : 0.906 : 1.002  : 1.041 : 1.002 :

• ~

• ~ 0 1.002  : 1.042  : 1.002  : 0.906  : 0.000  : 1.014 : 1.046 : 1.004 :

• ~

• ~

• : 1.019  : 0.952  : 1.015  : 1.002  : 1.014  : 1.030 : 0.937 : 1.022 :

• ~

1.032  : 0.974  : 0.952 : 1.041  : 1.046  : 0.937 : 0.979 : 1.034 :

1.000 : 1.032  : 1.019  : 1.002 : 1.004  : 1.022 : 1.034 : 1.002 :

FIGURE A.2 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)

( ~ )

ANF-90-01 Page A-7

• ~

• : 0.958  : 0.964  : 1.000  : 1.026  : 1.025 : 1.000 : 1.064 : 0.958 :

• ~

• : 0.964  : 0.981  : 1.049  : 0.922  : 1.031 : 1.047 : 1.019 : 0.964 :

• ~

• ~

• : 1.000  : 1.049  : 1.015  : 1.004  : 0.996 : 1.010 : 0.937 : 1.000 :

• ~

• ~

• : 1.026  : 0.922  : 1.004  : 0.000  : 0.937 : 0.995 : 1.031 : 1.026 :

• : 1.025  : 1.031  : 0.996  : 0.937  : 0.000 : 1.001 : 0.972 : 1.027 :

• ~

• ~

• : 1.000  : 1.047  ! 1.010  : 0.995  : 1.001 : 1.014 : 1.051 : 1.042 :

• ~

0.964  : 1.019  : 0.937  : 1.031  : 0.972 : 1.051 : 0.974 : 1.029 :

0.958  : 0.964  : 1.000  : 1.026  : 1.027 : 1.042 : 1.029 : 1.004 :

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

ANF-90-01 Page A-8

• ~

• : 0.967  : 0.969 : 0.997  : 1.019  : 1.019 : 0.996 : 0.968 : 0.966 :

• ~

• : 0.969  : 0.981  : 1.044  : 0.932 : 1.030  : 1.042 : 1.013 : 0.968 :

• ~

• : 0.997  : 1.044  : 1.017 : 1.008 : 1.001  : 1.012 : 0.944 : 0.997 :

• ~

• : 1.019  : 0.932  : 1.008  : 0.000 : 0.947  : 1.000 : 1.030 : 1.019 :

• ~

• ~

• : 1.019  : 1.030  : 1.001  : 0.947  : 0.000 : 1.006 : 0.976 : 1.020 :

• ~

• : 0.996  : 1.042  : 1.012  : 1.000  : 1.006 : 1.017 : 1.047 : 1.032 :

0.968  : 1.013  : 0.944  : 1.030 : 0.976  : 1.047 : 0.975 : 1.020 :

0.966  : 0.968  : 0.997  : 1.019 : 1.020  : 1.032 : 1.020 : 1.003 FIGURE A.4 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)

ia ANF-90-01 Page A-9

• ~

• : 0.984  : 0.976  : 0.995  : 1.012 : 1.012 : 0.995 : 0.976 : 0.984 :

• ~

• ~

• : 0.976  : 0.982  : 1.034 : 0.940  : 1.024 : 1.033 : 1.007 : 0.976 :

• ~

• ~

• : 0.995  : 1.034  : 1.015 : 1.008  : 1.002 : 1.011 : 0.950 : 0.995 :

• ~

• : 1.012  : 0.940 : 1.008  : 0.000 : 0.956 : 1.002 : 1.025 : 1.014 :

• ~

• ~

• : 1.012  : 1.024  : 1.002  : 0.956 : 0.000 : 1.007 : 0.981 : 1.015 :

• ~

• ~

• : 0.995  : .1.033  : 1.011  : 1.002 : 1.007 : 1.016 : 1.039 : 1.024 :

• ~

0.976  : 1.007  : 0.950  : 1.025 : 0.981 : 1.039 : 0.976 : 1.016 :

0.984  : 0.976  : 0.995  : 1.014 : 1.015 : 1.024 : 1.016 : 1.009 :

FIGURE A.5 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 CENTRAL FUEL)

I ep ANF-90-01 Page A-10

• : 0.984 : 0.976  : 0.995  : 1.012 : 1.012 : 0.995 : 0.976 : 0.984  :

• ~

• ~

• : 0.976 : 0.982  : 1.034  : 0.940 : 1.024 : 1.033 : 1.007 : 0.976  :

• ~

• : 0.995  : 1.034  : 1.015  : 1.008 : 1.002 : 1.011  : 0.950 : 0.995 :

• ~

• ~

• : 1.012 : 0.940  : 1.008 0.000 : 0.956 : 1.002  : 1.025 : 1.014 :

• ~

1.012  : 1.024  : 1.002  : 0.956 : 0.000 : 1.007  : 0.981 : 1.015 :

• ~

• : 0.995  : 1.033  : 1.011  : 1.002 : 1.007 : 1.016  : 1.039 : 1.024 :

• ~

0.976  : 1.007  : 0.950  : 1.025 : 0.981 : 1.039  : 0.976 : 1.016 :

0.984  : 0.976  : 0.995  : 1.014 : 1.015 : 1.024 : 1.016 : 1.009 :

FIGURE A.6 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 PERIPHERAL FUEL)

gl E

ANF-90-01 Page A-ll

• ~

• : 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 :

• ~

• ~

.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 :

1.03  : 1.00  : .99 : .99 : .99  : .99 : 1.00 : 1.03 :

FIGURE A.7 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL)

SO 80 70 60 Cl 50 C) 40 30 20 10 0

0 0.2 0.4 0.6 0.8 1.2 1.6 ECl R

lD I RRDIRL PONER PERKING PD lO I I FIGURE A.S RADIAL POWER HISTOGRAH FOR I/4 CORE SAFETY LIHIT HODEL

ANF-90-01 Issue Date: 1/10/90 WNP-2 CYCLE 6 PLANT TRANSIENT ANALYSIS Distribution:

D. J. Braun

0. C. Brown R. E. Collingham M. L. Hymas S. E. Jensen J. E. Krajicek S. L. Leonard J. L. Maryott L. A. Nielsen L. G. Riniker G. L. Ritter R. B. Stout/F.. B. Skogen H. E. Williamson Y. U. Fresk/WPPSS (51)

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