ML20141N181

Rev 1 to Washington Nuclear Power 2 Cycle 2 Plant Transient Analysis

ML20141N181
Person / Time
Site:	Columbia
Issue date:	02/28/1986
From:	Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17278A619	List: ML20141N181 ML20141N186
References
XN-NF-85-143, XN-NF-85-143-R01, XN-NF-85-143-R1, NUDOCS 8603040455
Download: ML20141N181 (50)

Text

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I g Xh-N 2-85-143 '

REVISIO\ 1 I

I W \ 3-2 CYC _ E 2 P_AN-~

I ~9A\SIE\~~ A\ A_YSIS I

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I TE39UARY 1986 I

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EXXO\ \ JC_ EAR COV 3ANY \C.

$03 $00$ 0 i7[

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XN-NF-85-143 Rev. 1 Issue Date: 2/24/86 i

WNP-2 CYCLE 2 PLANT TRAN51ENT ANALYSIS Prepared by:

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J.E. K $jicek l .

SWR Safety Analysis Concurt h VJZYue

' R.E. Collinglydim Manager 8WR Safety Ahalysis Concurs k k_ as -

L. Aorgan, Minager

//![d Custdeer Services Engineering Approve: M# 8/"//t I G.N. Ward, Manager Reload Licensing ed//. e el e w , ~

I Approve: v HDtiltamon, Manager

2. ' 1 %

Licensing & Safety Engineering Approve: Nd i bEEar',Managar'

/t 3/f t, Fuel Engir.eering & Technical Services min I

EXj(ON \UC_ EAR COV 3ANY, \jC.

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I NUCLEAR REGULATORY COMMIS$10N DISCLAIMER IMPORTANT NOTICE REGARDING CONTFNTS AND USE OF THIS DOCUMENT PLE ASE READ CAREFULLY This technical report was denved through research and development Il programs sponsored by Emmon Nuclear Company, Inc. It is being sty mitted by Emmon Nuclear to the USNRC as part of a technical contri-bunon to facilitate safety analyses by licensees of the USNRC whch utiliae Ennon Nuclear-fabricated reloa1 fuel or other techncal servmes prov=Jed by Ennon Nuciear for liant water power reactors and it is true and cortect to the best of Emmon Nuclear's knowledge, information, and behef, The information contained hereen may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are cusemers of Emmon Nuclear in their demonstration of comohance with the USNRC's regulations.

Without derogaang from the foregoing neither Emmon Nuclear nor any corson acting on its behalf. 3 A. Makes any warranty, empress or implied, with respect to the accur acy, completeness, or u

• fulness of the infor-mation contained in this documet, or that the use of any information, arcaratus, tv,ethod, or process disclosed in this document will not infnnge privately owned nghts.

or 8 Assumes any liabilities with respect so the use of, or for derr*Ms resulting from the use of, any information asy paratus, method, or process disclosed in th's document E

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I i XN-NF-85-143 Rev. 1 TABLE OF CONTENTS Section EDE

1.0 INTRODUCTION

........................................... 1 2.0

SUMMARY

................................................ 2 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN.................. 4 3.1 Design B3 sis........................................... 4 3.2 Anticipated Transients................................. 5 3.2.1 Load Rejection Without Bypass..........................

8 3.2.2 Feodwater Controller Fa11ure...........................

5 6

3.2.3 Loss of Feedwater Heating.............................. 7 I 3.3 3.4 Calculational Mode 1....................................

Safety Limit...........................................

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4.0 MAXIMUM 0VERPRESSURIZATION............................ 20 l

M 4.1 Design Bases.......................................... 20 4.2 Pressurization Transients............................. 20 4.3 Closure of All Main Steam Isolation Valves..... ....... 21 5.0 RECIRCULATION FLOW RUN-UP............................. 22

6.0 REFERENCES

............................................ 25 APPENDIX At MODIFICATIONS TO COTRANSA/PTSBWR3......... A-1 APPENDIX 8: COTRANSA HOT CHANNEL MODEL................ B-1 APPENDIX C: WNP-2 PLANf SPECIFIC FEATURES I INCORPORATED INTO COTRANSA/PTSBWR3........ C-1 APPENDIX 0: MCPR SAFETY LIMIT......................... D-1 8

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I XH-NF-85-143 e 11 Rev. 1 LIST OF TABLES Table h_qq 2.1 Thermal Nargin for Design Basis Conditions................t.. 3 i 3.1 Design Reactor and Plant Conditions for WNP-2................ 9 3.2 Significant Parameter Values Used in Analysis for WNP-2..... 10 3.3 Results of System Plant Transient Analyses. . . . . . . . . . . . . . . . . . 13 5.1 Reduced Flow NCPR Operating Limit for WNP-2. . . . . . . . . . . . . . . . . 23 I

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g 111 XN-NF-85-143 g Rev. 1 LIST OF FIGURES Fiqure I .

3.1 Load Rej ection Wi thout Bypass Resul ts. . . . . . . . . . . . . . . . . . . . . .

flie 14 3.2 Load Rejection Without Bypass Results...................... 15 3.3 Feedwater Controller Failurn I 3.4 Feedwater Controller Failure Results.......................

Results.......................

16 17 3.5 Loss of Feedwater Heating Results.......................... 13 3.6 Loss of Feedwater Heating Results.......................... 19 5.1 Reduced Flow MCPR Operating Limit.......................... 24 C.1 Thermal Power Moni tor Control System Logic. . . . . . . . . . . . . . . . . . C-5 C.2 Feedwater Control System.................................... C-6 C.3 Pre s s ure Re gul ato r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-7 0.1 WNP-2 Cycle 2 Safety Limit Local t

Peaking Factors (ENC Fue1).................................. D-5 D.2 WNP-2 Cycle 2 Safety limit Local Peaking Factors (GE Fuel)................................... D-6 i

D.3 Radial Power Histogram for 1/4 Core Safety Limit Model................................. D-7 I

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I I XN-NF-85-143 Rev. I

1.0 INTRODUCTION

This report presents the results of the Exxon Nuclear Company (ENC) evaluation of system transient events for the Supply System Nuclear Project Number 2 (WNP-2) during Cycle 2 operation. For this analysis the Cycle 2 core was assumed to contain 196 ENC 8x8 and 568 GE 8x8 fuel assemblies. The actual loading for Cycle 2 has not been defined at the i time the analysis was performed. However, the results presented herein are not expected to be significantly affected by changes in the number of fuel assemblies loaded.

I This evaluation together with core transient events (2) determines the necessary thermal margin (NCPR limits) to protect against the occur-

'4 rence of boiling transition during the most limiting anticipated transient. The evaluation also demonstrates the vessel integrity for

, the most limiting pressurization event. The core flow range has been expanded to include increased core flow up to the maximum attainable with the recirculation flow control valve in its fully open position.

The methodology for these analyses is detailed in Reference 1.

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I I 2 XN-NF-85-143 Rev. I 2.0.

SUMMARY

Using ENC methodology and considering Cycle 2 fuels, the most limiting I plant system transient for WNP-2 with regard to thermal margin was determined to be the generator load rejection without bypass (LRWB).

The Minimum Critical Power Ratios (MCPR) for potentially limiting plant system transient events at the design basis condition are shown in Table 2.1 for comparison. These transients were evaluated with all Cycle 2 co-resident fuel types medeled and the most limiting fuel type and exposure condition was used to determine the reported MCPR values.

The plant system transient MCPR values of Table 2.1 were obtained using a scram time based on WNP-2 measured values. The limiting transient (LRWB) was also analyzed at increased core flow conditions and a I. limiting delta CPR value of 0.19 (MCPR of 1.25) was obtained. The MCPR vaLas at the design basis and increased core flow analysis points were determined to be less limiting than those from the control rod with-drawal error (CRWE) core transient event. The CRWE analysis and resulting Cycle 2 MCPR operating limit are reported in Reference 2.

Maximum system pressure has been calculated for the containment isola-tion event, which is a rapid closure of all main steam isolation

. valves, using the scenario as specified by the ASME Pressure Vessel Code. This analysis shows that the safety valves of WNP-2 have suffi-cient capacity and performance to prevent the pressure from reaching the established transient pressure safety limit of 110% of design I, pressure. The maximum system pressures predicted during the event are below the ASME limit of 1375 psig(110%of design pressure) and are shown in Table 2.1. The analysis conservatively assumed six safety relief valves out of service.

The safety limit for all fuel types in Cycle 2 was determined to be 1.06 using the methodology of Reference 4.

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3 XN-NF-85-143 Rev. 1 Table 2.1 Thermal Margin For Design Basis Conditions Transient Delta CPR/MCPR Load Rejection Without Bypass 0.18/1.24 Feedwater Controller Failure 8

0.08/1.14 E.

Loss of Feedwater Heating 0.16/1.22 Maximum Pressure (osio)

Transient Vessel Dome Yessel Lower Plenum Steam Line MSIV Closure 1288 1317 1284 I

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i 104% power /100% flow.

Based on a safety limit MCPR 'ef 1.06.

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4 XN-NF-35-143

,I Rev. 1 3.0 TRANSIENT ANALYSIS FOR THERMAL MAJG.1H 3.1 Desian Basis Consistent with the FSAR plant transient analyses, transient analyses to determine the most limiting type of thermal margin transient were performed at the 104% power /100% flow Reload Analysis or Design Basis I point. The plant conditions used in the analysis for transients from this point are as shown in Table 3.1. The mest limiting transient was determined to be the generator load rejection without bypass. The most limiting exposure in cycle has been 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 determine thermal margin include ENC's I plant. transient and core thermal-hydraulic codes as described in deviousdocumentation(1,3-6) . Fuel pellet-to-clad gap conductances used in the analyses are based on calculations with RODEX2Y) . Recir-culation 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. Table 3.2 summarizes the values used for important parameters in the analysis.

Following requirements established in the Plant Operating License and associated Technical Specifications, and considering increased core flow, observance of a MCPR limit of 1.25 or greater conservatively protects against boiling transition during anticipated system plant I transients for WNP-2 Cycle 2. However, a Cycle 2 MCPR operating limit of greater than 1.25 is needed to protect the CRWE event as documented in Reference 2.

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5 XN-NF-85-143 Rev. 1 3.2 Anticioated Transients ENC considers eight categories of potential system transient occur-rences for Jet Pump BWRs in XN-NF-79-71(1) . The three most limiting transients are described here in detail to show the thermal margin for Cycle 2 of WNP-2. These transients are:

Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LOFH)

A summary of the transient analyses is shown in Table 3.3. Other plant E transient events are inherently nonlimiting or clearly bounded by one of the above events.

3.2.1 Load Reiection Without Bypass This event is the most limiting of the class of transients charac-

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terized by rapid vessel pressurization. The generator load rejection causes a turbine control valve trip, which initiates a reactor scram and a recirculating pump trip (RPT). The compression wave produced by the fast 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 conser-vatively 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 and 3.2 depict the time variance of critical reactor and ptant parameters from the analysis of the load rejection transient from the design basis point.

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I 6 XN-NF-85-143 Rev. 1 Analysis assumptions are:

I Control rod insertion time based on WNP-2 measured data Integral power to the hot channel was increased by 107.

consistent with Reference 8.

I-The LRWB was determined to be the limiting system transient; Table 3.3 shows that a delta CPR of 0.18 was calculated at the design basis l condition for the LRWB transient with RPT operable. In order to cover ,

the increased core flow conditions, the most limiting transient at the design basis condition (LRWB with RPT operable) was run at the 104/106Y.

power / flow point and a delta CPR of 0.19 was calculated.--

The load rejection was then analyzed assuming the same conditions but for the unusual case of the RPT and bypass inoperable. This resulted in a higher delta CPR for the RPT out of service, and in addition the delta CPR for the co-resident NSSS vendor fuel is slightly higher than the delta CPR for ENC fuel. The respective delta CPR values for the NSSS vendor and ENC fuels are 0.27 and 0.26 at the design basis condition and 0.28 and 0.27 for the increased core flow condition.

I 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 I no other action is taken. Eventually, the inventory of water in the downcomer will rise until the high vessel level setting is exceeded.

To protect against wet steam entering the turbine, the turbine trips I

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7 XN-NF-85-143 Rev. I 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 from the bypass valves opening. The evaluation of this event was performed using the scram and integral power assumptions discussed in 3.2.1 and 5 a

resulted in a delta CPR of 0.08 for the 104% power, 100% flow condition. Figures 3.3 and 3.4 present key variables for this event.

This event was also analyzed up to 104% power for increased core l flow operation. The results showed that FWCF transients from these conditions are also bounded by the control rod withdrawal error event as reported in Reference 2. Sensitivity results also show that the calculated delta . CPR is insensitive to the rate of feedwater flow increase.

3.2.3 Loss of Feedwater Heatina The loss of feedwater heating leads to a gradual increase in the subcooling of the water in the reactor lower plenum. Reactor power slowly rises to the thermal power monitor system trip setpoint.

The gradual power change allows fuel thermal response to maintain pace with the increase in neutron flux. For this analysis, it was assumed that the inicial feedwater temperature dropped 100 F linear-ly over a 50 second period. The magnitude of the void reactivity ,

feedback was assumed to be 10% lower than expected to assure that the power response to subcooling was gradual and the surface heat flux maintained pace with the power. Scram performance was based on WNP-2 data with scram worth 20% below expected. A delta CPR of 0.16 was E 3

calculated for the design basis condition as shown in Table 3.3; Figure 3.5 shows the key variables during this transient.

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I 8 XN-NF-85-143 Rev. 1 3.3 Calculational Model l

The plant transient code used to evaluate the pressurization transients I (generator load rejection and feedwater flow increase) was ENC's advanced code COTRANSA (I). This axial one-dimensional model predicted reactor power shifts toward the core middle and top as pressurization occurred. This was accounted for explicitly iri determining thermal B "^r9i" ch*"S*5 1" th* tr*"51*"t- ^ Pr'55"ri2*ti " tr""Si'"t5 **r*

B analyzed on a bounding basis using the COTRANSA hot channel delta CPR model as discussed in Appendix B. The loss of feedwater heating event was evaluated with the code PTSBWR3(I) since rapid pressurization and void collapse does not occur in this event. Appendix A delineates the I changes made to COTRANSA(I) to merge the PTSBWR3 code with the COTRANSA code, Appendix B describes the hot channel delta CPR code modification, and Appendix C covers the WNP-2 plant specific modeling for this application made to these codes for the WNP-2 analysis.

3.4 Safety Limit The 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 2 was determined by the methodology presented in Reference 4 to have a value of 1.06.

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

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9 XN-NF-85-143 Rev. 1 Table 3.1 Design Reactor and Plant Conditions I

For WNP-2 Reactor Thermal Power (104%) 3464 MWt Total Recirculating Flow (100%) 108.5 Mlb/hr Core Channel Flow 95.9 M1b/hr

] Core Bypass Flow 12.6 Mlb/hr Core Inlet Enthalpy 528.0 BTU /lbm Vessel Pressures g

Steam Dome 1034.7 psia 5 Upper Plenum 1049.4 psia Core 1057.4 psia Lower Plenum 1073.4 psia Turbine Pressure 989.7 psia g

Feedwater/ Steam Flow 14.98 Mlb/hr a Feedwater Enthalpy 403.8 BTU /lbm Recirculating Pump Flow (per pump) 16.28 Mlb/hr I

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I I 10 XN-NF-85-143 Rev. 1 I Table 3.2 Significant Parameter Values Used in Analysis For WNP-2 I High Neutron Flux Trip Void Reactivity Feedback 126.2 %

10% above nominal (I)

Time to Deenergized Pilot Scram Solenoid Valves 200 msec Time to Sense Fast Turbine Control Valve Closure 80 msec I Time from High Neutron Flux Time to Control Rod Notion 290 msec Scram Insertion Times (2) 0.404 see to Notch 45 0.660 sec to Notch 39 I 1.504 see to Notch 25 2.624 see to Notch 5 Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position T.*ip 90% open Turbine Control Valve Stroke Time (Total) 150 msec Fuel / Cladding Gap Conductance g Core Average (Constant) 595.5 BTU /hr-ft2-F S Safety / Relief Valve Performance Settings Technical Specifications Relief Valve Capacity 228.2 lbm/sec (1091 psig)

Pilot Operated Valve Delay / Stroke 400/100 msec I

I (I)For loss of feedwater heating.

I For rapid pressurization transients a 10% multiplier on integral is used; See Reference 8 for methodology description.

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

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11 XN-NF-85-143 Rev. 1 Table 3.2 Significant Parameter Values Used in Analysis For WNP-2 (Cont.)

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MSIV Stroke Time 3.0 sec l

MSIV Position Trip Setpoint 85% open Condenser Bypass Valve Performance l

Total Capacity 990.28 lbm/sec Delay to Opening (80% open) 300 msec Il

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Fraction of Energy Generated in Fuel 0.965 Vessel Water Level (above Separator Skirt)

High Level Trip (L8) 73 in Normal 49.5 in

' Cow Level Trip (L3) 21 in Maximum Feedwater Runout Flow ,

Two Pumps 5799.44 lbm/sec Doppler Reactivity Coefficient (I)

-0.00260 $/*F-void fraction Void Reactivity Coefficient (I)

-15.50 $/ void fraction Effective Delayed Neutron Fraction 0.00522 Prompt Neutron Lifetime 0.0479 msec Recirculating Pump Trip Setpoint 1170 psig Vessel Pressure -

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(I) Nominal value I

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I 12 XN-NF-85-143 Rev. 1 Table 3.2 Significant Parameter Values Used in Analysis For WNP-2 (Cont.)

I Control Characteristics I

Senior Time Constants I Steam Flow 1.0 see Pressure 500 msec Others 250 msec Feedwater Control Mode Three-Element I Feedwater 100% Mismatch Water Level Error 48 in Steam Flow Equiv. 100%

Flow Control Mode Manual Pressure Regulator Settings Lead 3.0 see Lag 7.0 sec Gain 3.33%/psid I

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I 13 XN-NF-85-143 Rev. 1 Table 3.3 Results of System Plant Transient Analyses

) I Maximum Maximum Maximum Core Average System Neutron Flux Heat Flux Pressure Event (% Rated) (% Rated) (psic) delta CPR Load Rejection I

f Without Bypass 320 115 1170 0.18

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Feedwater Controller Failure 147 109 1158 0.08 I

as Loss of Feedwater Heating 126 123 1069 0.16 MSIV Closure with Flux Scram 536 128 1317 N/A I

B Note: All results are for the bounding approach at the design analysis 104/100% Power / Flow point.

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I I 20 XN-NF-85-143 Rev. 1 4.0 MAXIMUM OVERPRESSURIZATION l Maximum system pressure has been calculated for the containment I. 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 I 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 pressur-ization event are those shown in Table 3.1. The most critical l active component (scram on MSIV closure) was assumed to fail I during the transient. The calculation was performed with ENC's ad-vanced plant simulator code COTRANSA(I) which includes an axial one-dimensional neutronics model.

4.2 Pressurization Transients ENC has evaluated several pressurization events and has determined that g closure of all main steam isolation valves (MSIVs) without direct scram E is the most limiting. Since the MSIVs 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 MSIVs are closed than when turbine valves are closed. The closure rate g of the MSIVs is substantially slower than the turbine stop valves or 5 turbine control valves. The impact of this smaller volume is more important to this event than the slower closure speed of the MSIV I

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21 XN-NF-95-143 Rev. I valves relative to turbine valves. Calculations base determined that the overall result is to cause MSIV closures to be more limiting than turbine isolations.

g 4.3 Closure of All Main Steam Isolation Valves This calculation also assumed that six relief valves were out of I

service and that all four main steam isolation valves were isclated at the containment boundary within 3 seconds. At about 3.3 seconds, the reactor scram is initiated by reaching the high flux trip setpoints.

Pressures reach the recirculating pump trip setpoint (1170 psig) before EI 3

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 maximum pressure calculated in the steam lines was 1284 psig occurring near the vessel at about 5 seconds.

gl The maximum vessel pressure was 1317 psig occurring in the lower plenum 5 at about 5 seconds. These results are presented in Table 2.1 and 3.3 for the design basis point. The MSIV overpressurization event was also analyzed at the increased flow condition and no significant difference l in the results (same within roundoff as values in Table 2.1) were l observed from the design basis point. '

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I 22 XN-NF-85-143 Rev. 1 5.0 RECIRCULATION FLOW RUN-UP The MCPR full flow operating limit is established through evaluation of I 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.

Exxon Nuclear determined the required reduced flow MCPR operating limit by evaluating a bounding slow ficw increase event. The calculations assume the event was initiated from the 104% rod line at minimum flow I and terminate 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 MCPR equals the safety limit at the final power / flow runup point. The reduced flow MCPRs were then calculated by XCOBRA(4) at discrete flow points. The reduced flow MCPR operating limit 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 operating limit and the full flow MCPR operating limit as summarized in Reference 2.

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23 XN-NF-85-143 Rev. 1 li W

Table 5.1 Reduced Flow MCPR Operating Limit i For WNP-2 l

Core Flow Reduced Flow MCPR

(% Rated) Operatino limit 100 1.07 90 1.12 80 1.17 l

70 1.23 60 1.32 '

50 1.42 ,

40 1.55 I

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l.0 20 30 40 50 60 70 80 90 100 tto O TOTAL CORE RECIRCULATING FLOW ( /'. RHTED)

RE00CEO FLOW MCPR OPERATING LIMIT I Figure 5.1 Reduced Flow MCPR Operat irit; 1.1mit i

I I 25 XN-NF-85-143 I

6.0 REFERENCES

1. R.H. Kelley, " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," XN-NF-79-71(P), Revision 2 (as supple-I mented) Exxor. Nuclear Co., Inc., Richland, WA 99352 1981.

November I 2. J.E. Krajicek, " Supply System Nuclear Project Number 2 (WNP-2)

Cycle 2 Reload Analysis," XN-NF-86-01, Revision 1. Exxon Nuclear Co., Inc., Richland, WA 99352, February 1986.

3. T.L. Krysinski and J.C. Chandler, " Exxon Nuclear Methodology for I Boiling Water Reactors: THERMEX Thermal Limits Methodology; Summary Description," XN-NF-PO-19(P), Volume 3. Revision 1 Exxon Nuclear Co., Inc., Richland, WA 99352, April 1981.

4 T.W. Patten, " Exxon Nuclear Critical Power Methodology for

. Boiling Water Reactors," XN-NF-524(P1, Revision 0. Exxon Nuclear Co., Inc., Richland, WA 99352, November 1979.

5. R.H. Kelley, "Dresden Unit 3 Cycle 8 Plant Transient Analysis Report " XN-NF-81-78, Revision 1, Exxon Nuclear Co., Inc.,

Richland, WA 99352, December 1981.

6. R.H. Kelley and N.F. Fausz, " Plant Transient Analysis for Dresden Unit 2. Cycle 9," XN-NF-82-84(P1, Exxon Nuclear Co., Inc.,

I 7.

Richland WA 99352, October 1982.

K.R. Merckx,"RODEX2 Fuel Rod Mechanical Recponse Evaluation I Model,* XN-NF-81-58(&l, Revision Richland, WA 99352, March 1984.

2 Exxon Nuclear Co., Inc.,

8. S.E. Jensen, " Exxon Nuclear Plant Transient Methodology for I Boiling Water Reactors: Revised Methodology for Including Code Uncertainties In Determining Operating Limits for Rapid Pres-surization Transients in BWRs " XN-NF-79-71(Pl. Revision 2 Supplement 3. Exxon Nuclear Co., Inc., Richland, WA 99352 March I 1985.

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I A-1 XN-NF-85-143 Rev. 1 APPENDIX A I NODIFICATIONS TO COTRANSA/PTSBWR3 COTRANSA originated with the coupling of a plant transient simulator code, PTSBWR3, and a one-dimensional, coupled neutronic hydraulic code, I COTRAN. Subsequent to the initial licensing of Dresden-2 with COTRANSA, the following modifications have been introduced into ENC's BWR plant transient model:

I o latest version of COTRAN replaced the original COTRAN I o control system input module (CONTROL) introduced to coding o codes COTRUI, COTRANSA, PTSBWR3, and CONTROL all reside in the same program library The latest version of COTRAN (JUL83) replaced the original COTRAN because the numerical convergence features had been upgraded to in-crease code execution efficiency. A Control System Nodule has replaced the original control system model so that all operations are handled through the input stream and may be easily tailored to the specific plant application. Having COTRAN, COTRANSA, PTSBWR3, and CONTROL in the same program library permits stand alone or grouped execution of I each of the codes.

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I I I I B-1 XN-NF-85-143 Rev. 1 APPENDIX B COTRANSA HOT CHANNEL MODEL B.1

SUMMARY

I The original COTRANSA hot channel model was used to give a figure of merit delta CPR for the determination .of the limiting transient. The limiting delta CPR was then obtained by the user through a XCOBRA-RODEX2-HUXY manual iteration. This involved a time consuming cal-culation where the user manipulated a considerable amount of data I between codes. Furthermore, it also resuited in a transient condition being analyzed using steady state approximations. The COTRANSA hot channel model has been modified to automate the delta CPR calculation and to give a transient delta CPR. Each fuel type is modeled, and a delta CPR specific to that fuel type is determined. XCOBRA and RODEX2 are used to dctermine the input for each hot channel. COTRANSA then calculates the delta CPR for each time step. The largest delta CPR is then reported.

B.2 MODIFICATIONS TO THE HOT CHANNEL I B.2.1 Flow Response Surface l The modifications to COTRANSA include a time dependent calculation of the flow rate to the hot assembly of each fuel type. The initial and transient flows to the hot channel are determined using XCOBRA. ENC's approved subchannel code for BWR's. A steady state response surface I for the hot assemblies' flow rates are determined for four key variables:

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B-2 XN-NF-85-143 Rev. 1 o Relative assembly thermal power o Core average thermal power o Core average active flow o Core pressure A quadratic equation is used to fit the above variables for the hot assembly flow for each fuel type.

B.2.2 Fuel Temoerature Model The fuel temperature model for the hot rod is consistent with the approved HUXY model as described in the Plant Transient Methodology for BWRs (XN-NF-79-71 Rev. 2) with the clad gap conductance based on RODEX2 calculations. Each fuel type is run to the end of cycle, at the end of cycle the power is increased and the relationship of gap conductance to rod average fuel temperature is then cetermined. This gap conductance is then used in the hot channel model.

B.2.3 Critical Power Ratio Calculations I

The MCPR calculation model used in the hot channel model is the approved XN-3 correlation as described in XN-NF-512 Rev. 1. The hot channel model calculations do not interact with the core average solutions since the impact of the hot assembly is so small. Therefore, the boundary conditions which drive the hot channel model are stored and used iteratively. These boundary conditions are:

o Power o Core inlet enthalpy o Pressure o Inlet flow rate o Outlet flow rate g) g I

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I I B-3 XN-NF-85-143 Rev. 1 o Bulk fluid temperature o Clad to fluid heat transfer coefficient o Heat flux I o Axial power distribution o Enthalpy rise I The purpose of the calculation is to determine the maximum allowable assembly power which will not exceed critical heat flux conditions I during the transient.

Since an iterative calculation is used to establish the initial power

{

in the hot channel, the initial power used in the calculation is only l I an estimate. After the completion of the transient simulation the lowest calculated CPR is compared to 1.0 and the power of the fuel rod is modified. This new power is assumed as an initial condition.

The flow to the limiting assembly is determined from the response surface and the enthalpy rise is ad. justed to be consistent with the new I conditions. The hot channel model calculations are repeated and the lowest CPR is again compared to 1.0. The process is then repeated until the lowest CPR is 1.0. The initial CPR minus the lowest CPR is the delta CPR for the transient consistent with ENC's reported methodology.

I B.3 VERIFICATION OF THE HOT CHANNEL  ?

~

Two different checks were made to insure the adequacy of the COTRANSA hot channel model's delta CPR calculation. The standard' XCOBRA -

RODEX2-HUXYiterationwasperforrgd,andsteadystatecoNditionswere input into the hot channel. The XCOBRA - RODEX2 - MJXY iteration I resulted in delta CPR's that were higher than those for the hot channel model. This is what was expected because of the tionservative . steady state nature of the hand method. When quasi-steady state conditions

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B-4 XN-NF-85-143 Rev. I were forced into the hot channel, the results of the comparison were improved and as expected. An XCOBRA-T verification analysis of the limiting transient confirmed the hot channel model result.

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I II C-1 XN-NF-85-143 Rev. 1 APPENDIX C HNP-2 PLANT SPECIFIC FEATURES INCORPORATED INTO CONTRANSA/PTSBWR3 C.1

SUMMARY

l The ENC plant transient analysis codes incorporate many plant specific features such as control systems through input and coding changes.

This appendix discusses the required WNP-2 plant specific changes and I control system input to the COTRANSA/PTSBWR3 plant simulator code.

These modifications were incorporated for exclusive use for WNP-2. l Unless otherwise stated, modifications that were made to the PTSBWR3 portion of the code were paralleled in the COTRANSA portion.

C.2 THERMAL POWER MONITOR The thermal power monitor (TPM) is designed to prevent reactor power increases above a conservative analytical limit of 120.0% of rated power, or, if operation is at off rated flow, then I Pscram = (0.66

• Wpct + 0.57)

I where Pscram = flow biased scram setpoint Wpct = recirculating flow through drive lines as % of rated I The measured power is intentionally time delayed with a six second time I constant to prevent spurious trips. If monitored power exceeds Pscram or 120% of rated power, the reactor scram system trips. The logic used in the plant transient simulator is in Figure C.I.

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C-2 XN-NF-85-143 Rev. 1 C.3 RECIRCULATING PUMP TRIP The recirculating pumps will automatically trip off on high vessel pressure or low downcomer level as determined by user input or, the user can default all RPT's. Additionally, the user may select an input option which would trip the pumps on indication of a turbine trip or a E 3

generator load rejection. When using this option, the user specifies the process rad mechanical delay time between the initiating event of j the pump trip (turbine trip or load rejection) and the time at which l the pump rotor begins to physically decelerate. The user als specifies by input the setpoints for tripping the pumps on vessel E' 5

pressure and level.

C.4 FEEDWATER CONTROLLER I '

Because of the modifications made to COTRANSA as discussed in Appendix A, the feedwater control system is incorporated into the code as part I

of the input stream. Figure C.2 gives a schematic of the control '

system used. '

The feedwater control system maintains a pre-established level in the reactor vessel during normal plant operation by varying the speed of the steam turbine driven feed pumps. Steam flow (WIV) and feed flow (WFW) are compared and an error signal is sent to the mismatch gain amplifier. The output from the mismatch gain amplifier is combined with the sensed water level (ZLEVEL) to provide a composite water level. After processed by a lead-lag controller, with time constants T1 and T2, to eliminate the signal fluctuation, the composite reactor water level is compared to the level setpoint (ZLEVSP) to provide the I

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I C-3 XN-NF-85-143 Rev. 1 input signal to the flow controller. This flow is controlled by a PI (proportional and integral) controller with constants of Kp and Ki.

I The function generator signal is then sent to the turbine feed pump. A lag function (Tp) is used to model the pump inertial effect.

C.5 RECIRCULATION FLOW CONTROL SYSTEM I WNP-2 is a base loaded plant, so the recirculation flow control system I will always be in manual control. No fl,w control system was modeled and the valve position was manually controlled through the code initialization.

C.6 PRESSURE REGULATOR CONTROL SYSTEM The pressure regulator control system is entered into COTRANSA/PTSBWR3 as input data. The model as it is input is shown in Figure C.3.

Functionally, the pressure regulator adjusts turbine and bypass flow to maintain turbine pressure at a desired setpoint. Essentially, the system produces an error signal by comparing a sensed pressure (PTT) with a pressure setpoint (PTTTT). This error signal is conditioned by the lead / lag characteristics of the control valve and produces a steam flow based on the pressure setpoint.

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C-4 XN-NF-85-143 Rev. 1 I'

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APRM TPM Trip --- s TPM \

1 TP _ 4 >4 SR APRM :

1+T TP S TP = 6 sec d Scram I

SE Orive  ; /

riow I

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Figure C.1 Thermal Power Monitor Control System Logic I

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I D-1 XN-NF-85-143 Rev. 1 APPENDIX D HCPR SAFETY LIMIT D.1 INTRODUCTION I 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 l 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 I 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 I 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.

I The MCPR safety limit is established through statistical consideration of measurement and calculational uncertainties associ ; 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 corre-lation, power peaking, and core coolant distribution, are fuel related.

I When ENC fuel is introduced into a core where it will reside with another supplier's fuel types, the appropriate value of the MCPR safety limit is calculated based on fuel-dependent parameters associated with I

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I-I D-2 XN-NF-85-143 Rev. I the mixed core. Similarly, when an ENC-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 MCPR safety limit is again reevaluated. g i ll The design basis power distribution is made up of components cor-responding to representative radial, axial, and local peaking factors.

Where such data are appropriately available from previous cycles, these factors are determined through examination of operating data for previous cycles E and predictions of operating conditions during the 5 cycle being evaluated for the MCPR safety limit. If operating data are not available, either because the reactor has not been operated or because appropriate data cannot be supplied to ENC, the safety limit power distribution is determined strictly from the predicted operating conditions during the cycle being evaluated. Data for WNP-2 are limited to Cycle 1 operation, which is not considered typical of later cycle operation because of the differences in fuel exposure distributions as well as between G.E. initial core fuel and both ENC and G.E. reload fuel designs.

D.2 ASSUMPTIONS D.2 Desian Basis Power Distribution I

The local, radial and axial power distributions which were determined to be conservative for use in the safety limit analysis are shown in Figures D-1, through D-3.

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I D-3 XN-NF-85-143 Rev. I D.2 HYDRAULIC DEMAND CURVE Hydraulic demand curves based on calculations with XCOBRA were used in I the safety limit analysis. The XCOBRA calculation is described in ENC 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."

D.2.3 SYSTEM UNCERTAINTIES System measurement uncertainties are not fuel dependent. The values reported by the NSSS supplier for these parameters remain valid for I the insertion of ENC fuel. The values used in the safety limit analysis are tabulated in ENC topical report XN-NF-524(A), "Exyon

(

Nuclear Critical Power Methodology for Boiling Water Reactors."

D.2.4 FUEL RELATED UNCERTAINTIES I Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty. The values used in the safety limit analysis are tabulated in ENC topical report XN-NF-524(A), " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."

I Power measurement uncertainties are established in EHC topical report XH-NF-80-19(A), Volume 1. " Exxon Nuclear Methodology for Boiling Water Reactors; Neutronics Methods for Design and Analysis."

D.3 SAFETY LIMIT CALCULATION I A statistical analysis for the number of fuel rods in boiling I transition was performed using the methodology described in ENC topical reportXN-NF-524(A), " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors." After 500 Monte Carlo trials it was I

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D-4 XN-NF-85-143 E Rev. 1 5 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%.

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I D-5 XN-NF-85-143 Rev. 1 I

I LL L ML M M ML L LL 0.91 0.95 1.02 1.06 1.06 1.02 0.95 0.91 I L 0.95 ML 0.97 H

1.08 ML*

0.87 H

1.04 H

1.07 M

1.04 L

0.95 ML H H H H H ML* ML 1.02 1.08 1.01 1.00 0.98 1.00 0.90 1.02 I

M ML' H W M H H M l 1.06 0.87 1.00 0.00 0.90 0.98 1.04 1.06 M H H M W H M M 1.06 1.04 0.98 0.90 0.00 0.99 0.93 1.05 I ML H H H H H H M 1.02 1.07 1.00 0.98 0.99 1.00 1.07 1.08 I L 0.95 M

1.04 ML' 0.90 H

1.04 M

0.93 H

1.07 ML' 0.96 ML 1.07 I LL L ML M M M ML L 0.91 0.95 1.02 1.0S 1.05 1.08 1.07 1.03 Figure 0.1 WNP-2 Cycle 2 Safety Limit x u u s24 Local Power Factor (ENC Fuel)

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0-6 XN-NF-85-143 Rev. 1 I

LL L ML M M ML i

L i LL Il 1.03 1.00 0.99 0.99 0.99 0.99 1.00 1.03 I

L M H H MH MH ML L a 1.00 0.99 1.03 1.02 0.99 0.99 0.97 1.00 g ML H L' H H MH MH ML 0.99 1.03 0.91 1.02 1.01 0.98 0.99 0.99 M W H H MH M I

H H 0.99 1.03 1.02 0.00 1.02 1.01 0.99 0.99 M

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Figure D.2 WNP-2 Cycle 2 Safety Limit Local *""

l Peaking Factors (G. E. Fuel)

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WNP-2 CYCLE 2 DESIGN BRSIS RRDIRL POWER IG -

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g XN-NF-85-143 W Rev. 1 Issue Date 2/24/86 WNP-2 CYCLE 2 PLANT TRANSIENT ANALYSIS I

, Distribution J.C. Chandler I R.E. Collingham S.E. Jensen

! T.H. Keheley l J.E. Krajicek

'g T.L. Krysinski

'W J.L. Naryott J.N. Norgan G.L. Ritter G.N. Ward H.E. Williamson

, J.B. Edgar /WPPSS (32) l Document Control (5)

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