ML17199U413

Cycle 11 Plant Transient Analysis

ML17199U413
Person / Time
Site:	Dresden
Issue date:	09/30/1987
From:	Braun D SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
References
ANF-87-096, ANF-87-96, NUDOCS 8803150339
Download: ML17199U413 (67)

Text

ANF-87-096 ADVANCED NUCLEAR FUELS CORPORATION DRESDEN UNIT 3 CYCLE 11 PLANT TRANSIENT ANALYSIS.

SEPTEMBER 1'987

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AN AFFILIATE OF KRAFTWERK UNION QKwu 8803150339 880309 1 '

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ADVANCED NUCLEAR FUELS CORPORATION Prepared By:

DRESDEN UNIT 3 CYCLE 11 PLANT TRANSIENT ANALYSIS D. J. Braun ANF-87-096 Issue Date: 9/28/87 BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services.

AN AFFILIATE OF KRAFTWERK UNION

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f CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-cerning the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document is issued. Accordingly. except as otherwise expressly pro-vided in such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the infor*

matlOn contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned

.. rights: or assumes any llabilitles with respect to the use of any information, ap-paratus. method or process disclosed in this document.

The lnformatlOn contained herein Is for the sole use of Customer.

In order to avoid lmpaltment of rights of Advanced Nuclear Fuels Corporation in

• patents or inventions which may be included in the information contained in this document. the *recipient, by its acceptance of this document, agrees not to publish or make.Public use (in the patent use of the term) of such information until so authorized in writing by Advanced Nuclear Fuels Corporation or until after six (8) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the fumishing of this docu-ment.

XN-NF*F00-765 (1167)

1

- i -

ANF-87-096 Section 1.0 2.0 3.0 3.1 3.*2 3.3 3.3.1 3.3.2 3.3.3 3.3.4

. 4 4.1 4.2 4.3 5.0 5.1 5.2 5.2.1 5.2.2 6.0 TABLE OF CONTENTS INTRODUCTION......................................................

SUMMARY

TRANSIENT ANALYSIS FOR THERMAL MARGIN...........*....*.*.*........

Design Basis........*.***.*****.**......*..*..*.......***.........

Calculational Model...*.**.**** *....*....*..*..*...................

Anticipated* Transients *..***.*.....*.....*.....*...*..............

Load Rejection Without Bypass *****.............*....*.............

Feedwater Controller Failure....................*.............. :~*.

Loss Of Feedwater Heating............................... *~*........

Statistical Uncertainty Analysis............

L *********************

MCPR Safety limit..................................................

MAXIMUM OVERPRESSURIZATION ANALYSIS~..............................

Design Basis......................................................

Pressurization Transients.........................................

Closure Of All Main Steam Isolation Valves........................

ANALYSIS AT OFF-RATED CONDITIONS..................................

Increased Core Flow and Reduced Feedwater Temperature.............

Pump Run up Event *...*****..*........*.**...*.....*..**............

Automatic Flow Control..********..*..***.***.***.*................

Manua 1 Fl ow Contra 1..*...*.*..****.....*...**.*..**...*....*......

REFERENCES........*.**.........*......*...*.*.....................

APPENDIX A - SINGLE LOOP OPERATION Page 1

3 8

8 8

9 9

10 11 12 13 32 32 32 33 37 37 37 38 39 51 A. l ABNORMAL OPERATING TRANSIENTS..................................... A-1 A.1.1 Load Rejection Without Bypass..................................... A-2 A.1.2 Feedwater Controller Failure...................................... A-2 A.1.3 Pump Seizure Accident......................... -.................... A-3 A.1.4 MCPR Safety Limit................................................. A-4 A.1.5 Summary........................................................... A-4 A.2 MAPLHGR LIMITS.................................................... A-5 A.3 STABILITY......................................................... A-6

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ANF-87-096 LIST OF TABLES 2.1 Rated Power Delta CPR's - Dresden Unit 3 Cycle 11.................

5 2.2 Thermal Margin Summary - Dresden Unit 3 Cycle 11..................

6 2.3 Results Of Plant Transient Analyses - Dresden Unit 3 Cycle 11.....

7 3.1 Design Reactor And Plant Conditions - Dresden Unit 3.............. 15 3.2 Significant Parameter Values Used In Analysis - Dresden Unit 3.... 16 3.3 Control Characteristics.......................................... ~

18 3.4 Data Used In Statistical Transient Analysis....................... 19 3.5 Input For MCPR Safety Limit Analysis............................... 20 5.1 5.2

.3

-~ 5.4 Coastdown Transient Analysis Results.............................. 41 ASME Overpressure Results......................................... 42 Automatic Flow Control Excursion Path............................. 43 Reduced Flow MCPR Limits For Automatic Flow Control (8x8 Fuel).... 44 5.5 Reduced Flow MCPR Limits For Automatic Flow Control (9x9 Fuel)~... 45 5~6 Manual Flow Control Excursion Path................................ 46 5.7 Reduced Flow MCPR Limits For Manual Flow Control................ ~.

47 A.I SLO Reactor And Plant Conditions.. ~******************************* A-7

- iii -

ANF-87-096 LIST OF FIGURES Figure

1 ANF-87-096

1.0 INTRODUCTION

This report describes the plant transient analyses performed *by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 11 (XN-4) reload for Dresden Unit 3.

Cyc 1 e 11 is the fourth eye 1 e during which ANF fue 1 will be irradiated in Dresden Unit 3.

The Cycle 11 core is to contain both ANF 9x9 assemblies and ANF 8x8 assemblies. Operating limit critical power ratio values of these fuel types for plant system transients during Cycle 11 operation are established in this report.

The analyses reported in this document were performed using the same average core plant transient analysis methodology as was used to calculate thermal argin requirements for Cycle 10 of Dresden Unit.3 (Ref. 1 and 2).

The approved XCOBRA-T hot channe 1 mode 1 was used to ca lcu 1 ate the 1 i mit i ng delta CPR's.

This analysis supports operation in the expanded power/fl ow operating map

• shown in Figure 1.1.

Secti.on 5.0 describes the results of the off-r:ated

• analysis performed to demonstrate that the MCPR operating 1 imits, together with reduced flow MCPR, allow operation throughout this map. *

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Figure 1.1 Dresden 3 Proposed Operating Power/Flow Map

• 3 ANF-87-096 2.0

SUMMARY

The determination of thermal margin requirements for Dresden Unit 3 Cycle 11 was based orr the consideration of various operational transients.

The limiting transients in each general category of event are i dent ifi ed in Reference 2.

The most limiting transient events for determination of thermal margins in BWR/3 applications were determined to be the generator load rejection without bypass to the condenser, the loss of feedwater heating event, and the feedwater controller failure (maximum demand) event.

For the case of Dresden Unit 3, the most limiting of these events was found to be the generator load rejection without bypass.

Table 2.1 presents the change in critical power ratio (delta CPR) at bounding conditions for the three most limiting transients.

ifhe 'Safety Limit for Cycle 11 nominal (two loop operation) conditions was calculated to be 1.05; this value is applicable to all fuel types.

The minimum thermal margin MCPR operating limit for each fuel type in Cycle 11 of Dresden Unit 3 is contained in Table 2.2. These are obtained by adding the delta CPR's of the limiting transient in Table.2.1 to the 1.05 Safety Limit.

These results are applicable for rated and increased core flow conditions.

Maximum system pressure for ASME overpressure evaluation has been calculated for the postulated closure of all main steam isolation valves (MSIV's) without activation of the MSIV position scram and without pressure relief credit for the four electromatic relief valves.

The results of this analy~is as shown in Table 2.2 indicate that the requirements of the ASME Code regarding overpressure prot~ction will continue to be met for the Dresden Unit 3 Cycle 11 core; the calculated pressures are below the 1375 psig limit.

iTransient analysis results at off-rated conditions are presented in Section 5.0.

The results confirm that the reduced flow MCPR limits of Cycle 10 remain

4 ANF applicable i'n Cycle 11, and th~t the full power MCPR operating limit covers the off-rated conditions of reduced feedwater temperature and reduced power.

Results of the SLO analysis are shown in Appendix A.

The safety limit analysis for single loop operation supports an increase of.01:

5 TABLE 2.1

• RATED.POWER DELTA CPR'S

• .DRESDEN UNIT 3 CYCLE 11.

Delta CPR Transient ANF 8x8 Generator Load Rejection Without Bypass(!)

Feedwater Flow C92troller Failure tMaximum Demand)l J Loss of Feedwater Heating(2)

(!)Delta CPR on statistical basis.

0.23 0.20 0.18

( 2 l Delta CPR based on Technical Spec if i ca ti on scram performance.

ANF-87-096 ANF 9x9 0.26 0.23 0.19

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Transient TABLE 2.2 THERMAL MARGIN

SUMMARY

DRESDEN UNIT 3 CYCLE 11

. MCPR Operating* Limit(!)

Generator Load Rejection Without Bypass ANF 8x8 1.28 Maximum Pressurization (psiq)

Transient MSIV Closure Without Position Scram (ASME)

Steam Dome 1297

(!)Based on a 1.05 safety limit.

Lower Plenum 1324 ANF-87-*

ANF 9x9

1. 31 1297

• 7 TABLE 2.3 RESULTS OF PLANT TRANSIENT ANALYSES DRESDEN UNIT 3 CYCLE 11 Maximum Maximum Core Average Neutron Flux Heat Flux Event

% of Rated

% of Rated Generator Load Rejection Without Bypass(!)

319 115 Feedwater Flow Controller Failure (Maxi.mum Demand) 241 117 oss-- of Feedwater eating 120 119 MSIV Closure (ASME Analysis) 439 134

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(!)Nominal conditions; all other events are for bounding conditions.

ANF-87-096 Maximum Vessel Pressure (psiq) 1259

.:4-1184 1034 1324

8 ANF-87-096 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN This section describes the analyses which were performed to determine the MCPR operating limits for Cycle 11 of Dresden Unit 3 at nominal conditions (full power and core flows between 100 and 108%).

3.1 Design Basis The* plant transient analyses for Dresden* Unit 3 determined that the most limiting condition for thermal margin at reactor operation at full power was increased core flow (ICF), 108%.

Reactor plant conditions for these analyses are shown in Table 3.L The most limiting point in the cycle is when the control rods are fully withdrawn from the core.

The thermal margins es tab 1 i shed for the end of full power capabi 1 ity are conservative for cases here control rods are partially inserted.

Observance of the MCPR operating limits shown in Table 2.2 will provide adequate protection against the occurrence of boiling transition during all anticipated transients for Cycle 11 operation of Dresden Unit 3 at nominal conditions.

3.2 Calculational Model The average core plant transient methodology described in References 2 and 4 as updated in Appendix A of Reference 1 was used for the analysis reported in this document.

The approved XCOBRA-T(S) hot channel model was used to calculate the delta.CPR's rather than the COTRANSA hot channel delta CPR model.

• The COTRANSA one-dimensional core model is used to mo_del the axial power shifts associated with the system overpressurization in the generator_ load rejection and feedwater controller failure transients.

The PTSBWR point

.kinetics core model is used for the loss-of-feedwater heating transient.

Fuel pellet-to-cladding gap conductance values used in the analyses were based on RODEX2 calculations for the Dresden Unit 3 Cycle 11 core configuration.

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9 ANF-87-*

In accordance with ANF methodology, possible limiting.transients are evaluated using a consistent set of bounding input.

From these bounding results, the limiting transient is identified as the generator load rejection. without bypass.

Since this is a rapid pressurization event, ANF's methodology for including uncertainties in determining operating limits for rapid pressurization transients in BWRs (Ref. 8) is used.

This methodology includes code uncertainties and uncertainties in important input variables.

A conservative deterministic integral power multiplier of 110% accounts for code uncertainties when* the statistical methodology is being applied.

This code uncertainty methodology was previously-use<;I for Dresden. 3 Cycle 10 (Ref. 1).

. Table 3.2 summarizes the values used for important parameters in the analysis.

Table. 3.3 provides the feedwater flow, recirculating coo.lant flow, and pressure regulation system settiAgs ~sed in th~ analysis.

3.~

Anticipated Transients Eight major categories of transients were considered generically in Reference

2.

For Cycle 11 operation*_Qf Dresden Unit 3, specific events have been evaluated for thermal margin.

These events are the generator load rejection transient without bypass to the condenser, the feedwater controller failure to maximum demand, and loss of"*,:feedwater heating.

In the analyses, it was

• assumed that a relief valve was out-of-service. Reference 2 demonstrated that other categories of transients *are. either. inherently sel f-1 imiting or bounded by one of these.

3.3.1 Load Rejection Without Bypass

.The Load Rejection Without Bypass (LRWB) is the most limiting of the rapid pressurization transients and of all the system transients for Dresden Unit 3.

This conclusion has been verified through comparison with the results of the analysis of the turbine trip transient without turbine bypass for the Ores.

plants.

=..,

10 ANF-87-096 In* the load. rejection transient, steam flow is interrupted by an abrupt closure of the turbine control valve.

The resulting pressure increase causes a decrease in the void level in the core, which in turn creates a power excursion.

This excursion is mitigated in part by *Doppler br,oadeni ng

• and pressure relief, but the primary mechanisms for termination of the event are control rod insertion and revoiding.

The important parameters for this transient include the power transient (integral power) determined by the void reactivity,. which affects the initial power excursion rate and part of the intrinsic shutdown mechanism, and the control rod worth,. whi.cn.:.determines the value of the scram reactivity.* Other important inputs include: the control rod movement parameters (scram del 1ay and insertion speed),* *which2 determine the event characteristics following the initial mitigation of the power excursion.

The bounding case at full power onsidering ICF conditions resulted in delta_ CPR's of 0.30 and 0.33 for the ANF 8x8 and the ANF 9x9 fuels respectively as shown in Table 5.1.

3.3.2

. Feedwater Controller Failure Failure of the feedwate'r control.system could lead to a maximum incr~ase of feedwater flow into ttie reactor vessel; The excessive feedwater *flow increases the subcooling**in the** recirculating water returning to the reactor core.

This reduction in average moderator temperature will result in the core's power rising to attain a new equilibrium if no other action.is taken.

Eventually, the level of water in the downcomer region will rise until the high water level trip is reached., The turbine then trips to. prevent the transmission of liquid water to the turbine, and the turbine stop valves

. close.

The resulting scram arrests the power increase, and the pressure pulse resulting from the stop valve closure is suppressed by the opening of the bypass valves.

The *analysis assumed that all of the conservative conditions of Table 3;2 were.concurrent; the calculated delta CPR is*~ bounding.result.

The calculated values for the transient, as shown in Tables 2.1 and 5.1, are

11 ANF-87-*

0.20 for the 8x8 fuel and 0.23 for the 9x9 fuel. These values are bounded by the values for the LRWB event.

Figures 3.1 and 3.2 illustrate the behavior of major system variables during the Feedwater Controller Failure transient.

3.3.3 Loss Of Feedwater Heating The loss of feedwater heating leads to a gradual increase in the subcooling of the water in the lower* plenum.

Core power slowly rises to the*overpower trip setpoint.

The gradual power change allows the fuel thermal response to maintain pace with the increase in neutron flux.

In this analysis, it is conservatively assumed that the feedwater temperature dropped 200 degrees F over a two-minute period.

Vaid reactivity is assumed to be 25% more negative than the nominal calculated value, which resulted in a maximum value of po and heat flux.

Scram performance is assumed to be at Technical Specificatio limits, and control rod worth is assumed to be 20% less than the nominal calculated value.

These conservatisms are.sufficient. to cover cycles with these fuel types such that the:~esults can be referenced for future cycles.

Previous loss of feedwater heating analyses (including Ref. 1) for the Dresden Units have shown that the *delta CPR of the transient to. be less limiting than the*.above* trahsients.

Furthermore, the delta CPR for the transient is expected to be approximately the same for 8x8 and 9x9 fuel types because it is a slow transient and thus, rod thermal response times are unimportant.

This analysis again confirmed that this transient is not limiting for any fuel type in Cycle 11 of Dresden Unit 3.

The result of the loss of feedwater heating transient analysis is a maximum delta CPR of 0.19 for the ANF 9x9 fuel type as shown in Tables 2.1 and 5.1.

Figures 3.3 and 3.4 illustrate the behavior of major system variables duri~

the Loss of Feedwater Heating transient.

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12 ANF-87-096 3.3.4 Statistical Uncertainty Analysis The bounding transient analysis showed the load rejection without bypass (LRWB) to be the liriliti.ng transient for Dresden Unit.3 Cycle.11; as in previous cycles.

When rapid pressurization transients are limiting, ANF methodology for including plant and code uncertainties in the determination of MCPR operating limits may be applied (Ref. 4 and 8). This methodology uses a conservative deterministic multiplier of 110%

on the calculated power transient and treats the uncertainties in the important input variables (scram speed and scram delay) statistically.

The delta CPR's used to. establish *the MCPR operating limits result from the use of the deterministic 110% integral power combined with 95% probability that the stat i st i ca 11 y determi netl

• delta CPR is not exceeded.

The uncertainty in the control rod drive performance parameters, scram delay, and insertion speed was determined _.from measured pl ant data.

lncorporat i ng the most recent plant data, the uncertainty in the scram del~y time for Cycle 11 was determined to correspond* to.*:~ mean value of 241 msec and a stan(iard.

deviation of 29 msec. *in the.Cycle JO analysis of Unit *3, a mean value~~.of 241 msec and a standard deviation of 28 msec were used.

Similarly, the uncertajnty in the-*scram insertion speed for Cycle 11 *was determined to.. correspond to a mean value... of 137.. 9 cm/sec and a. standard deviation of 3.6 cm/sec.

In the Cycle 10 analysis, a mean of 139.6. cm/sec and a standard deviation of 3.3 cm/sec were used.

The uncertainties for the Cycle 11 analyses are compared with the uncertainties for the Cycle 10 analyses (Ref. 1). in Table 3.4.

Figures 3.5 and 3.6 illustrate the behavior of major system.variables. during the LRWB transient using nominal input for uncertainty variables.

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i3 ANF 3.4.

MCPR Safety Limit The MCPR safety 1 imit for Cycle 11 operation of Dresden Unit 3 was determined using the methodology described in References 1 and 6.

This methodology was used to determine the MCPR safety limit for Cycle 10 operation of Dre~den Unit 3 (Ref. 1).

As with the transient delta CPR's, the Dresden 3 Cycle 11 safety limit analysis assumed constant flow (no flow iteration) in computation of delta CPR.

This is consistent with the Dresden 3 Cycle 10 analyses (Ref. 1).

The main input parameters and uncertainties used in the safety limit analysis are listed in Table 3.5.

The de~ign basis radial ~ewer distribution used in the analysis is shown in Figure 3.7.

This power distribution is found to have the greatest number rods near limits for Cycle 11 and is conservative in this regard relative other expected power distributions during the cycle.

The most conservati radial power distribution determined for Cycle 11 is modified by moving the radiali for' four maximum pow~~ assemblies* to a radial peaking f~ctor corresponding to a MCPR near the expected MCPR operating 1 imit.

Using the radial. power distribution in Figure 3.7 is expected to bound Cycle 11 operation and.will support the established safety limit. 'Five fuel types were represented in the Dresden 3 Cycle 11 safety limit analysis; i.e., ANF XN-4 9x9, ANF XN-3 9x9, ANF XN-2 8x8, ANF XN-1 8x8 centr~l ~uel, and ANF XN-1 8x8 peripheral fuel.

I Bounding local power distributions for each fuel type over their expected Dresden 3 Cycle 11 exposures were used.

The local power distributions for the ANF fuel types are shown in Figures 3.8 and 3.11.

A flat local power distribution (all locals equal unity) was conservatively used for ANF XN-1 8x8 peripheral fuel bundles.

The results of the analysis support a two 1 oop MCPR safety 1 imit of 1. 0.

Protection of this limit will assure that at least 99.9% of the fuel rods in the core are expected to avoid boiling transition during normal operation and

14 ANF-87-096 anticipated operational occurrences.

This limit. applies *to all of the fuel types in Cycle 11.

. 15 ANF TABLE 3.1 DESIGN REACTOR AND PLANT CONDITIONS:

DRESDEN UN IT 3 Reactor Thermal Power Total Recirculating Flow Core Channel Flow Core Bypass Flow Core Inlet Enthalpy Vessel Pressures Steam Dome Upper Plenum Core Lower Plenum Turbine Pressure Feedwater/Steam Flow Feedwater Enthalpy Recirculating Pump Flow (per pump) 2527 MWt 98.0 Mlb/hr

88. 2 Ml b/hr

• 9.8 Mlb/hr 522.3 Bt!J/lbm 1020 psia 1026 psia 1035 psia 1049 psia 964.7 psia 9.8 Mlb/hr 320.6 Btu/lbm 17.1 Mlb/hr

.* 16 ANF-87-096 TABLE 3*~2. SIGNIFICANT PARAMETER VALUES USED IN ANALYSis(l)

DRESDEN UNIT 3 High Neutron Flux Trip Control Rod Insertion Time Control Rod Worth Void Reactivity Feedback Time to Deenergized Pilot Scram Solenoid Valves Time to Sense Fast Turbine Control Valve Closure Time from High Neutron Flux Trip to Control Rod Motion ttTurbine Stop Valve Stroke Time Turbine Stop Valve Position Trip Turbine Control Valve Stroke Time (Total)

Fuel/Cladding Gap Conductance Core Average (Constant)

Safety/Relief Valve Performance Settings Pilot Safety/Relief Valve Capacity Power Relief Valves Capacity Safety Valves Capacity Pilot Operated Valve Delay/Stroke Power Operated Valves Delay/Stroke 3032.4 MW 3.5 sec/90% inserted 20% below nomina1(2) 25% more neg. than nomina1(2) 283 msec (maximum) 80 msec 290 msec 100 msec 90% open 150 msec 614 Btu/hr-ft2-F Technical Specifications 166.1 lbm/sec (1080 psig) 620.0 lbm/sec (1120 psig) 1432.0 lbm/sec (1240 psig) 400/100 msec 967/200 msec (l)LRWB transient was evaluated statistically (see 3.3.4)

(2)used in Loss of Feedwater Heating Transient.

17 ANF TABLE 3. 2 Sl(iNIFICANT PARAMETER VALUES USED IN ANALYS.IS DRESDEN UNIT 3 (Continued)

MSIV Stroke Time 3.0 sec MSIV Position Trip Setpoint Condenser Bypass Valve Performance Total Capacity Delay to Opening (from demand)

Opening Time (entire bank, max demand)

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

Norma*l Range of Operation High Level Trip Maximum Feedwater Runout Flow Three pumps Two pumps 90% open 1085.2 lbm/sec 100 msec 1.0 sec 0.965 30 in 20-40 in 48 in 4966 lbm/sec 3310.67 lbm/sec

-0.00231 $/°F

• Doppler Reactivity Coefficient(!)

Void Reactivity Coefficient(!)

Effective Delayed Neutron Fra~tion Prompt Neutron Lifetime Recirculating Pump Trip Setpoint

-16.55 $/void fraction

(!)Nominal value.

0.0051152 0.04502 msec 1240 psig Vessel Pressure

18

.ANF-87-096 TABLE 3 ~ 3

• CONTROL CHARACTERISTICS Sensor Time Constants Pressure 100 msec Others 250 msec.

Feedwater Control Mode One-Element Feedwater Master Controller Proportional Band 100%

Reset

5. rep/min

~*J, Feedwater~iOO% Mismatch I~

Water.tevel Error 60 in Steam Flow (not used) 12 in eq.

Flow Control Mode Master Manual Master Flow Control Settings Proportional Band 200%

Reset 8 rep/min Speed Controller Settings Prop?;,tional Band 350%

Reset 20 rep/min Pressure *setpoint Adjuster Over a 11 Gain 5 psi/% demand Time Constant 15 sec Pressure Regulator Settings Lead 1.0 sec Lag 6.0 sec Gain 3.33 %/psid

19 ANF TABLE 3.4 DATA USED IN.STATISTICAL TRANSIENT ANALYSIS

. C:tcl e 10 C:tcle 11 Variable Mean Std Dev Mean Std Dev Scram Insertion Speed (cm/sec) 139.6 3.3 137.9 3.6 Scram Delay Time (msec) 241 28 241

-29

20 ANF-87-096 TABLE 3;5 INPUT.FOR MCPR,SAFETY LIMIT ANALYSIS.

Fuel-Related Uncertainties Uncertainty Statistical

• Parameter Source Document Percent Nominal Treatment XN-3 Correlation XN-NF-512(A) 4.11 Convoluted

. XN-NF-734(A)

Radial Peaking Factor XN-NF-80-19(A) 5.28 Convoluted Volume 1 Local Peaking Factor XN-NF-80-19(A) 2.46 Convoluted Volume 1 Axial Peaking Factor XN-NF-80-19(A) 2.99 Limiting Value Volume 1 Assembly Flowrate XN-NF-79-59(A) 2.80 Convoluted Plant Measurement Uncertainties Uncertainty

• Stat i st i cal Parameter Units Value Percent

. Treatment Feedwater Flowrate Mlbm/hr 12.4*

1. 76 Convoluted Feedwater Temperature deg F 340.1 0.76 Convoluted Core Pressure psi a 1035 0.50 Convoluted Total Core Flow Mlbm/hr 98.0 2.50 Convoluted Core Inlet Temperature 2.00 Replaced by core inlet enthalpy Core Inlet Enthalpy Btu/lbm 522.3 2.40 Convoluted Core Power MW 3200*

All owed to vary with heat balance

• Feedwater flowrate and core power were increased above design values to attain desired core MCPR for safety limit evaluation, consistent with XN-NF-524(A), Revision 1 (Ref. 6).

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t:o

z,,

I °'

-....J I

0 \\0

°'

c::::J LlJ 600 500 400

~300 LL.

0

...... z

~200 a:

LlJ a..

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.II

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~

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HEA r FLUX 3.. REC RCULATI lN FLOW

4.

VES 3EL STEA ~ FLOW

l.

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3 3

3 w

c:

2 L 1.--v-4 4

/-

1 1.

1 J

-10. u.O 0.5 1.0 1.5 2.0 2.5 3.0 3.5 TIME, SEC Figure 3.5 Generator Load R on Without Bypass 3

~

2

~4 -

1 1

4.0 4.5 3

2 4 ~

5.0 N

(.Tl

):>

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5.0 N

. °'

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'-I I

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0. 6

o. 8 1

1. 2

1. 4

1. 6

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-;-.i RAD!

OWER PEAKING

~

Figure 3.7 Design asis*Radial Power Distribution

28 ANF-87-096 LL LL L

L ML Ml L

l LL

0.97 : 0.92 :

1.00 ' : 0.97 : l.05 : 1.05.: 0.96 : 0.98 : 0.90 :

LL L

ML M

M

Ml*

M

Ml*

. L

0.92 ':' 0.98 :

1.02 : l.ll : l.08 0.93 :

l.10 : 0.99 : 0.95 :

. ------~---~----~-~-------------~-----------------~---~~-----~---------------------

L.

• Ml * :

• ML*

M H

H M

M ML 1.00 :

J.02 ' : 0.94 1.02 : J.06 : 1.06 ' : 1.00 : 1.05. : 0.99 :

~ ************p**~~--~---~--~*************~*****************************************

'* *W** *:

L

. ::M.. :

M H'

H H

H RI.**

*

• M

0.97 : l.11-* :***l.02* :

1.04 : 1.01

0.99 :

l.01

0.85 : 1.05 :,,,

ML M

H H

W :

H.

'H H

H 1.05 :

1.08 :

1.06 : 1.01

0.00 : 0.98 : 0.98 : 1.02 : 1.10 :

---

~--~~---~-------------------------------------------------------------

ML

*ML*

H

'! **fl

-H W

H ff H

1.05 : 0.93 :

l.06 : 0.99 : 0.98 : 0.00 : 0.98 : J.01

1.09 L

.M.

M H

H H

H

llll*.

M

• : 0.96 :. l_-:~O :

l.00 :

1.01

0.98. : 0.98 :

LOO

0,83 : 1.03

• --------~----------------------~-~*--~-~--~--------------------~-------------------

L ML*

ML* * * :

• H H

ML*

H

• ML

o.98 : o.99 :

1.05 : *o.s5 * :, 1.02 : 1.01

o.83

1.os : o.9s :

---

**-~------------------------------------------------

~

LL L

Ill

.: 1' * * -:.:.11 H

M * *

"1.

L

'; '0.90 ; 0.95 ; '0.99 '; '1.05 ~:;>'.1.10 ; 1.09 ;

1.03 ; 0.95 ; 0.91

.~. :- :.** *:*..

Figure 3.8 Design Basis Local Power Distribution For D3 9x9 3.35-9Gd4.0

29 ANF-87-0 LL LL L

l ML Ml l

L LL

0.97

0.93 : 0.97

0.96 I.OZ l.01

0.95 : 0.96 : 0.91.:

LL L

ML* *:

M M.

ML*

M Ml

** L

• : 0.93

0.96 : 0.99 :

1.08 :

1.07

0.94

l.07 : 0.98 : 0.94 :

L

Ml*

Ml M

H H

M M

ML

0.97 :. 0.99 : 0.97 1.03 1.07 1.06 : l.01

1.04 : 0.97

. :. --------------~----------------------------------------------------------------~--

L M

M H

H H

H

ML*

• M

0.96 :

1.08 : 1.03 :

1.05 :

1.03 1.0Z

1.03 : 0.88 : 1.04' :

• Ml: M: H: H:

W: H: H: H: H:

I.OZ 1.07 :

1.07 1.03 0.00 1.01 1.00 : 1.03

1.08 :

• ML*

ML*

H H

• H W

H H

H 1.01

0. 94

1.06 :

I.OZ l.01

0.00 :

1.00 : 1.03 : 1.08 :

L

. M

• :

• M

• H H

H H

ML*

M.

• . : 0.95 : :1.07 :

l.01 1.03 1.00 :

1.00 : 1.01

0,87

1.03

---

~---------------------------------------------

L Ml M

.. ML*

H H

ML*

H Ml

0.96 : 0.98 :

1.04

0.88 :. 1.03 1.03 : 0.87 1.07 : 0.96 LL l

ML

. M

. H H

M Ml L

0.91

0.94 :

0.97 :

1.04 :. l.08 :

1.08 :

l.03 : 0.96 : 0.92 Figure 3.9 Design Basis Local Power Distribution For D3 9x9 3.35-8Gd4.0

30 ANF-87-096 I

LL L

ML ML Ml Ml ML L

1.04 1.02 1.01 1.00 : 0.99

1.00. :

1.01 1.01 L

ML ML*

M M

M Ml*

ML

• : 1.02 1.00 : 0.95 : 1.05 :

1.04 :

1.04 : 0.94

0.99 :

ML ML*

M.. :

H H

H M

ML

• : 1.01 0.95

. 1.04 :

1'.03 1.02 1.02 :

1.02 :

0.96 :

Ml M :

H * :

H H

H H

R 1.00... :

1.05 :. 1.03 * :

Loo : 1.00 :

I.oo :

1.oi

• 1.03 *:

Ml M

H H

W H

H M

• : 0.99 1.04 1.02 1.00 : 0.00 : 0.99 :

1.00 :

1.02 Ml M

H H

H H

Ml*

M 1.00 :

}.04 :

1.02 : 1.00 : 0.99 : 0.99

0.88 :

1.02 ML

. ML*

M H

.H ML*

M ML 1.01 0.94 1.02 : 1.01

1.00* : 0.88 :

1.01

• =

0:95 L

Ml Ml M

M M

Ml ML 1.01 0.99 : 0.96 : 1.03 :

1.02 : 1.02

0.95. :

0.98 Figure 3.10 Design Basis Local Power Distribution For D3 8x8 3.02-6Gd3.0'

31 LL l

ML ML ML ML ML L

1.05 :

l.02 :

1.01 1.00 : 0.99 : 1.00. :

1.01 1.02

---

~------------------------------------------------------------

L

ML*

ML M

M M

ML*

Ml 1.02 : 0.97

0.97

1.04 :

1.03 : 1.04 : 0.95 : 0.99 :

Ml Ml M*

H :

H :

H :

M Ml 1.01 0.97 1.03 : 1.02 :. 1.01 1.02 : 1.02 : 0.96 :

Ml M. :

~

ff H * :

H :

H :

M : **

1.00 :

1.04 1.02 :

1.00 : 0.99 : 0.99 :

1.01 1.02 ML M

H H

W H

H M

0.99 1.03 1.01

0.99 : 0.00 : 0.99 : 1.00 :

L02

. ------------------------------------------------------------------------~

Ml M

H H

H H

Ml*

M 1.00 :

I.04 1.02 : o.99 :

o.~9 : o.99 * : o.88 : 1.02

• . ---------------------------------------------~---------------------------.

Ml ML*

M :

H H

ML*

M :

ML 1.01

0.95 :

1.02 :

l.01

1.00 : 0.88 : *1.01

0.96 :

---

~---------------------------------------

L ML ML M

M M

ML ML 1.02

0.99 : 0.96 : 1.02. : '1.02 : 1.02

0.96 : 0.98 :

• ------------------------------.---~------------------------------~-------

Figure 3.11 Design Basis Local Power Dist~ibution For 03 8x8 2.87-5Gd3.0 ANF-87-0.

32 ANF-87-096 4.0 MAXIMUM OVERPRESSURIZATION ANALYSIS This section describes the analysis of the maximum overpressurization accident performed for compliance with the ASME code.

4.1 Design Basis The reactor conditions used in the evaluation of the maximum pressurization transient are summarized in Table 3.1. These conditions are the same as those used in the transient analyses for thermal margin.

In addition to these conservative assumptions,. furth~r conservatism was added. by disallowing the operation of the four power-actuated relief valves as required by th*e ASME code.

Failure of the most critical active component was assumed.

In this instance; the most critical active component is the reactor trip associated with the.position of the Main Steam Isolation Valves (MSIV's).

4.2 Pressurization Transients Based on earlier analyses (Ref. 7), it has been determined that the maximum pressurization transient for the Dresden plants *is the inadvertent cl-0sQre of all MSIV's with failure of the MSIV.position scram.

The position scram, which commands reactor shutdown* almost* immediately upon MSIV move.ment, mitigates the effects of this event to the point that

• H does not contribute to the determination of thermal margins.

Delaying the scram until the high pressure trip setpoint is reached results in a substantially more severe transient.

Although the closure rate o~ the MSIV's is substantially slower than that of the turbine stop or control valves, the compressibility of the fluid in the steam lines provides significant damping of the compression wave associated with the turbine.trip ~vents to the point that the slower *MSIV closure without direct scram results in nearly as severe a compression wave.

Once the containment is isolated,. the subsequent core power production must be contained within a smaller system volume than that associated with the turbine

33 ANF-87-0.

trip events.

Comparative analyses hav.e demonstrated that the containment isolation event under these conservative assumptions results in a higher overpressure than either the turbine trip or the generator load rejection without bypass.

4.3

• Closure Of All Main Steam Isolation Valves This calculation assumed that all four steam lines were isolated at the containment boundary within three seconds.

The valve characteristics and steam compressibility combine to delay the arrival of the compression wave at the core until approximately three seconds from the initiation of the MSIV stroke. Effective shutdown is delayed until approximately 5 seconds following initiation of the MSIV stroke because control rod performance is assumed to be at the Technical Specification limit~.

As presented in Section 5.1, the most limiting condition for this transie was at ICF and normal feedwater temperature.

At about 3.1 seconds, the reactor scram is initiated by reaching the high flux trip setpoints.

Since scram performance was degraded to its Technical Specification limit, effective power shutdown is delayed until after 3.5 seconds.

Substantial thermal power production enhances pressurization.

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 maximum vessel pressure (~t the lower plenum) of 1324 psig was observed at approximately 6.4 seconds.

The maximum steam *line pressure of 1297 psig was also observed at approximately 6.4 seconds.

The maximum value of the sensed pressure in the steam dome was 1297. psig.

The relative values of maximum pressure during the MSIV closure transient indicate that the vessel and steam lines will be protected against overpressure limits defined in the ASME Co~

if a* pressure safety limit of 1375 psig in the lower plenum is protected.

34 ANF-87-096 Figures 4.1 and 4.2 illustrate the performance of major system* variables during the overpressuri zat ion accident.

This calculation was performed with COTRANSA.

See Section 5.0 for MSIV closure analysis at "off-rated" conditions.

't'

CJ LU l-600 500 400

~300 IJ..

0 1-z

~200 a:

LU Cl.

100 4

.... 3 *., *.... 3 *.,

4

.... 3 J

- 1'0.o 1.0 2.0

1.

Nl:.U ~UN YC0, l:EVl:.L

2.

HEA FLUX

3.

REC.RCULATI ~N FLOW 4.

VES ~EL STEA ~ FLOW

5.

• FEt: JWATER F.uw

~

\\

\\ J\\

}

2 "l 2 3.-

-* ';i

"":::....__ 2 3 s:

t:.7

i

i

~

\\

\\

4 4

\\~

~

~~ _p ~

1 v

3.0 4.0 5.0 6.0 7.0 TIME, SEC Figure 4.1 MSI sure 5

5 c

4 1

1 8.0 9.0 3

i::

4 5

~

10.0 w

01

)::>

2.,,

I

())

I 0

300 250 200 150 100 50 1

-5'1 u.O 2

2 2

1 1

1. 0 2.0

l.

Vt:::I >i=-L r-nc.::i ~

ll~C.

lt":l.L' 27

>EL WATE (IN)

L

_1 1

I I

l J I

~

r----~

__/

.___g_

2

~

3.0 4.0 5.0 6.0 7.0 8.0 9.0 TIME. SEC Figure 4.2 MSIV Closure 10.0

)>

z

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CX>

I 0

l.O C'I

37 ANF-87-096 5.0 ANALYSIS AT OFF-RATED CONDITIONS Transient analysis of a BWR requires consideration of transients at "off-rated" conditions.

This section describes those analyses performed in support of Cycle 11 that were not covered in Sections 3.0 and 4.0.

The generator load rejection without bypass and feedwater controller failure to maximum demand have been evaluated at three reduced flow points on the expanded power to flow map of Figur.e 1.1 and the results reported in Reference 1.

The analysis indicated that all three of these points are bounded by the 100% power/100%

. flow position on the operating map.

5.1 Increased Core Flow and Reduced Feedwater Temperature For Cycle 11 operation analyses were also performed for Increased Core Flow and for a feedwater temperature reduction of l00°F.

These analyses were performed at the end. of the operating cycle where all rods are fully withdrawn.

The result~ of the thermal margin analyses are summarized in Table 5.1.

Comparison of the results within Table 5.1 indicates that ICF had an insignificant effe.ct on the LRWB and FWCF delta CPR's.

Reduced feedwater temperature had a small effect on delta CPR's for LRWB and FWCF at 100% power.

The delta CPR's were larger at 108% ICF.

A corresponding ~aximum overpressure event was also analyzed.

Consistent with the nomi na 1 conditions, the most 1 imit i ng event was assumed to be a 11 four main steam isolation valves shutting within 3. seconds.

The results are summarized in Table 5.2.

Comparison of the results within Table 5.2 shows that the ICF at normal feedwater temperature is the limiting condition.

The results are discussed in Section 4.3.

5.2 Pump Runup Event Analysis for pump runup events for operation at less than rated recirculation pump capacity indicates* the need for an augmentation of the full fl ow MCPR

\\

38 ANF-87-0 operating limit for lower flow conditio~s.

This is due t6 the potential for large reactor power increases should an uncontrolled pump flow increase occur.

The. present *analysis establishes. the necessary reduced flow MCPR operating limit to protect the reactor fuel against.boiling transition during anticipated pump runup events from off-rated core flow conditions for both automatic flow control and manual flow control.

These limits are shown in Figures 5.1 and 5.2 for the aut6matic flo~ control event.

Figure 5.3 details MCPR limits pertaining to the manual flow control event.

The cycle specific MCPR limit for Dresden Unit 3 shall be the maximum of the reduced flow MCPR operating limit depicted in these figures for the appropriate fuel type and control mode and full flow cycle specific MCPR operating limit.

. 5. 2.1 Automatic Flow Control If the reactor is operated in the automatic flow control mode (AFC),

variations in core ~o~er shou~~ not result in critical power ratios less than the established MCPR operating* limit for rated conditions.

If the rated condition MCPR 1 imit is observed in a reduced flow condition, a subsequent increase in power to full power along the AFC control line may result in inadvertent degradation o! fuef critical power ratios to below this reference (full power) MCPR operating 1 imit.

The probability of boiling transition conditions occurring during a subsequent anticipated event may increase beyond acceptable levels if this were the case.*

ANF has determined the required reduced flow MCPR operating limit for off-rated conditions to prevent the MCPR from degrading below the Cycle 11 MCPR (full fl ow) operating 1 imits during AFC operation.

This was determined by evaluating the MCPR for a given reactor power distribution at varying total reactor power and flow conditions.

The variations in total core power

• flow were assumed to follow the expected relationship (Table 5.3) automatic flow control operation (100% rod line).

The power distribution chosen was such that MCPR equaled the referenced MCPR operating limit at rated

39 ANF-87-096 conditions of power. and flow.

The expected,variation of core pressure~ and inlet coolant subcooling with reactor power level was also considered.

I Additional results are presented for an adjusted power distribution such that MCPR at rated conditions of power and fl ow was increased by 0. 04 and by 0. 08.

The reduced flow MCPR limits were confirmed by

~COBRA (Ref. 6) calculations along the 100% rod line for AFC and are presented graphically in Figures 5.1 (8x8 fuel) and 5.2 (9x9 fuel), and in tabular form in Tables 5.4 (8x8 fuel) and 5.5 (9x9 fuel).

5.2.2 Manual Flow Control This section discusses pump excursions when the plant is not in automatic flow control operation mode, i.e., manual flow control.

Based on the results btained from previous Dresden Unit *a*nalyses (Ref. 7) which showed two pump excursions were the limiting pump !unup event, only two pump excursions are evaluated for Dresden Unit 3 Cycle 11.

The analysis of the two pump flow excursion indicates that the limiting event scenario is a gradual quasi-steady runup due to the inlet enthalpy lag associated with a more rapid runup.~ These results indicate that MCPR would decrease below the safety limit if th'e full flow reference MCPR was observed af initial conditions.

Thus, an augmented MCPR is needed for partial flow operation to protect the two pump excursion event.

The power to fl ow path used for the runup is shown in Table 5. 6 and bounds that calculated by XTGBWR for the constant Xenon assumption.

The reduced. fl ow MCPR ca lcul at ions have been performed assuming the event was initiated from the APRM Rod Block Line as well as the 100% flow control line.

The results show that pump runup events initiated from the minimum pump speed flow point on the 100% flow *control line to the 120% power/110% flow point in a straight ine relationship are bounding.

40 ANF-87-*

The results of *the t~o pump runup an~lyses* for

~anual flow control are presented in Figure 5.3 and Table 5.7.

The cycle specific MCPR limit for Dresden Unit 3 shall be the maximum of the reduced flow MCPR.operating limit or the full flow MCPR.operating limit.

41 ANF-87-096 TABLE 5.1 COASTDOWN TRANSIENT ANALYSIS RESULTS Power Flow ACPR Transient

__l_

~

ANF 8x8

• ANF 9x9 *

• LRWB 100 100

.29

.32 100 108

.30

.33 80 108

.30

.34 60 108

.31

.35 40 108

.28

.31 FWCF 100 100

.19

.21 100 108

.20

.23 80 108

.20

.23 60 108

.12

.14 40 108

.05

.06 40 100

.05

.06 LRWB w/FHOOS*

100 108

.27

.30 FWCF w/FHOOS*

100 108

.19

.21 80 108

.20

.23 60 108

.08

.09 40 108

.04

.04

• Feedwater Heater Out Of Service.

Note:

All transients analyzed with Technical Specification Scram Speeds.

~*

42 ANF. TABLE 5. 2. ASME OVERPRESSURE RESULTS Maximum Pressure Power Flow Vessel Dome Steam Line

~

~

psig Rii9.

psig 100.

108 1324 1297 1297

• 100 108 1304 1277 1279 100 87 1324 1303 1303

• Feedwater Heater Out Of.Service.

.43 ANF-87-096 TABLE 5.3 AUTOMATIC*FLOW.CONTROL.EXCURSION PATH Recirculating Flow Power

(% Rated)

• (% Rated) 100 100 90 94 80 88 70 81 60 74 50

. 67 40 58

44 TAbLE 5.4. REDUCED FLOW MCPR ~IMITS FOR AUTb~~TIC FLOW CONTROL (8X8 FUEL)

Recirculating Flow

{% Rated}

MCPR Limit 100 1.281 1.322 90 1.31 1.35 80 1.34 1.39 70 1.39 1.44 60 1.45 1.49

50.

1.52 1.56 40 1.66

1. 71 l1f MCPR Operating Limit = 1.28 at Rated Conditions.

~If MCPR Oper~ting limit = 1.32 at Rated Conditions.

If MCPR Operating Limit = 1.36 at Rated Conditions.

1.363 1.39 1.43 1.48 1.54 1.61

1. 76 ANF *

~~:. -

-~-

~~:*.

45

,TABLE 5.5 REDUCED FLOW MCPR LIMITS.FOR.AUTOMATIC FLOW CONTROL (9X9 FUEL)

Recirculating Flow

(% Rated)

MCPR Limit 100 1.311 1.352 90 1.34 1.38 80 1.36 1.41 70 1.41 1.45 60 1.46

1. 51 50 1.53 1.58 40 1.68

1. 73 lJf MCPR Operating Limit = 1.31 at Rated Conditions.

2rf MCPR Operating Limit = 1.35 at Rated Conditions.

31f MCPR Operating Limit = 1.39 at Rated Conditions.

1.393 1.42 1.45 L49 1.55 1.62

1. 78 ANF-87-096

. 46 ANF TABLE 5. 6 MANUAL FLOW CONTROL EXCURSION PATH **

Recirculating Flow Power

(% Rated}

• (% Rated) 110 120 100 111 90 102 80 93 70 85 60 76 50 67 40 58

47 ANF-87-096 TABLE 5.J REDUCED FLOW MCPR LIMITS FOR.MANUA~ FLOW CONTROL Recirculating Flow MCPR Limit

(% Rated) 8x8 9x9 100 1.10 1.09 90 1.15 1.14 80

1. 21 1.20 70 1.28

1. 26 60 1.36 1.34 50 1.46 1.44 40 1.60 1.57

2.0 MCPR OPERATING LIMIT = 1.28 1.9

---

MCPR OPERATING LIMIT = 1.32 E-t

• • • • • • • • • • • MCPR OPERATING LIMIT - 1.36

ill 1.8

~

p::

~

1.7

\\..

u

ill

~

1.6 0

~

rz..

1.5 C:l rzl u

14 C:l rzl

~

13 1.2........ -~

...... --..--.....---..--~--...---....--...... --..... ---.

20 30 40 50 80 70 80.

90 100 110 120 TOTAL CORE RECIRCULATING FLOW (% RATED. 98 MLB/HR)

Figure 5.1 Reduced Flow MCP Auto Flow Control (8x8 Fuel)

~

CX>

)::>

z I

CX>

'-J I

E--4 t-4

El t-4

~

p::

'14 u

-~

P=

0

~

. rz..

~

rz1 u

~

rz1 p::

2.0---------------------------------------------------------

1.9 1.8 1.7 1.6 1.5 1.4 1.3

\\..

\\.

\\.. '.. '..

--- MCPR OPERATING LIMIT -

1.31

---

MCPR OPERATING LIMIT -

1.35

• * * * * * * * * *

• MCPR OPERATING LIM IT = 1.39

~

1.2------------------------------......

20 30 40 50 60 70 80 90 100 110 120 TOTAL CORE RECIRCULATING FLOW.(% RATED, 98 MLB/HR)

Figure 5.2 Reduced Flow MCPR For Auto Flow Control (9x9 Fuel)

)>

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"° °'

1.7

. 8X8 FUELS E-4 1.8

---

9X9 FUELS

\\

\\

~

\\

1.5

' \\

~

\\

ll..

\\

\\

u

~ 1.4

~

0

~ 1.3 rz..

~

rz1 u

1.2

~

rz1

~

1.1 1.0-...-------------------------------------------------------......

20 30 40 50 80 70 80 90 100 110 120 TOTAL CORE RECIRCULATING FLOW (%*RATED. 98 MLB/HR)

Figure 5.3 Reduced Flow For Manual Flow Control t.n.

0

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z

-ri I

CX>

I

51 ANF-87-096

6.0 REFERENCES

1.

T. H. Keheley, "Dresden Unit 3 Cycle 10 Plant Analysis, XN-NF-85-62, September 1985.

2.

R.

H. Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-7l(P), Revision 2 (as supplemented),

Exxon Nuclear Co., Inc., Richland, WA 99352 (November 1981).

3.

K. R. Merckx, "RODEX2. Ftlel Rod Thermal Mechanical Response Evaluation Model," XN-NF-81-58(A), Revision 2, Exxon Nuclear Co., Inc., Richland, WA 99352 (March 1984).

4.

J. A. White, "Exxon Nuclear Methodology for Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary Description,"

XN-NF 19(P)(A), Volume 3, Revision 2, Exxon Nuclear Co., I.nc., Richland, WA 99352 (January 1987).

5.

"XCOBRA-T:

A Computer Code for BWR Transient Therma 1-Hydraul i c Core Analysis," XN-NF-84-105(P), Volume 1 and Supplements 1 and *2, June 1985.

R. B. Macduff and N. F. Fausz, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors," XN-NF-524(P)(A), Revision 1, Exxon Nuclear Co., Inc., Richland, WA 99352 (November 1983).

7.

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).

8.

S.

E.

Jensen, "Revised Methodology for Including Uncertainties in Determining Operating Limits for Rapid Pressurization Transients in BWRs," XN-NF-79-71(P), Revision 2, Supplement 3~ Exxon Nuclear Co., Inc.,

Richland, WA 99352 (March 1985).

l,

A-1 ANF-87-096 APPENDIX A SINGLE LOOP OPERATION The NSSS supplier has provided analyses which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended period of time.

These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed when both reci rcul at ion systems are in operation.

The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power.

Because the ANF fuel was designed to be compatible with the co-resident fuel in thermal hydraulic, nuclear, and mechanical design performam:e, and because the ANF methodology has given results which are consistent with those of the previous analyses for normal two-loop operation, the analyses performed by the NSSS supplier for single loop operation are also app.licable to single loop operation with fuel and ~nalyses provided by ANF.

A discussion of the relevant events and limits for single loop operation follow.

Also included are results of ANF analyses which confirm the NSSS vendor conclusions.

A.I ABNORMAL OPERATING TRANSIENTS MCPR limits established for full flow two loop operatton are conservative for single loop transients because of the physical phenomena related to part-power part-fl ow operation, not because of features in reactor analysis models or

• compatible fuel designs.

A review of ~he most limiting delta CPR transients for single loop operation was conducted.

Under single loop conditions, steady state operation cannot exceed approximately 81% power and 58% core flow because of the capability of the reci rcul at ion loop pump.

Thus, the MCPR limit at maximum power is higher than the two pump operating MCPR limit due to the fl ow dependent MCPR function.

This fl ow dependence is based on a fl ow

A-2 ANF-87-0.

increase transient from runup of two pumps.

Flow ru~ups from a single recirculation pump would be much less severe, though the conservative two *pump limit is retained.

A.I.I Load Rejection Without Bypass The limiting system transient for the Dresden Units is the Load Rejection Without Bypass (LRWB) pressurization transient.

In this transient, the primary phenomena is the pressurization caused by abruptly stopping the steam flow through rapid closure of the turbine control valve.

When the rapid press~ri zat ion reaches the core it causes a power excursion due to void collapse.

The reduced power and flow analyses for the Dresden Units described in Reference I.under two-loop conditions show that the resulting power excursi and associated delta CPR are reduced below those of the full power/full fl case.

Thus for the Dresden Units the MCPR 1 imits based on LRWB analyses at full-.power are conservatively applicable to the lower powers/flows associated with.single lOop conditions.

Furthermore, LRWB analyses by ANF at reduced powe_r: and flow conditions in other BWR's with single loop operation confirm this trend.

A.1.2 Feedwater Controller Failure The second most limiting transient for Dresden is the Feedwater Controller.

Failure (FWCF).

This transient is also less severe at the reduced power and flow conditions associated with single loop operati~n.

This transient assumes the feedwater controller fails to maximum demand and a 11 ows the ma~d mum amount of subcoo 1 ed feedwater into the down comer.

When this cooler water reaches the core, the power rises.

The core power rise ~

terminated through a turbine trip scram initiated by a high water level tri"B"'

in the downcomer due to the additional amount of feedwater being injected.

i:...

A-3 ANF-87-096 At the reduced re~irculation flows the subc~oling in the downcomer due to high

• feedwater injection takes longer to transverse to the ~ore such that a high 1eve1 trip occurs before the core power rises as much as in the full power case.

In the subsequent pressurization transient, the result of turbine trip is less severe for the reduced powers *in transients from single loop conditions because of the reasons discussed in the LRWB transient.

Thus,* because of the slower transport phenomena caused by the 1 ower fl ow in the downcomer and because of the lower steam line flow in the pressurization portion of the transient, and the higher full-power MCPR limit; the FWCF has

• 1arge~ margin *to :*th.e operating limit in single loop operation_ than in full-power two ioop operation.

The off-rated analyses in Section 5.0 of this report confirm that the FWCF delta CPR's for reduced power, eveQ at 108%, are less severe than the delta CPR's associated with the MCPR operating limit.

A.1.3 Pump Seizure Accident Pump sei2ure is a p6stulated atcident where the operating recirculation pump suddenly stops rotating.

This causes. a rapid decrease in core ftow, a decrease in the rate at which heat can be transferred from the fuel rods and a decrease in the cri ti ca 1 power ratio.

Ana 1 yses with COTRANSA and XCOBRA-T show the MCPR for ANF fuel would decrease by 0.43 during a pump seizure from single loop operation.

The COTRANSA code was used to simulate system response *ta a pump seizure in single loop operation *from the conditions specified* in Table A.I..

• The operating recirculation pump rota~ was *stopped in 0.1 seconds causing*a sudden decrease in active jet pump drive flow..

At about 1.9 seconds the inactive jet pump diffuser flow went from negative flow to positive flow.

The plant scrams

• on high water level at 2.1 seconds.

In 2.3 seconds the dome pressure

.ecreased to a minimum value _of 986 psia and then started to increase again.

Figures A.I and A.2 present a graphical representation of important system para.meters during the transient.

A-4 ANF-87-0.

The delta CPR for this event was calculated using XCOBRA-T.

The ANF 9x9 fuel reached a maximum delta CPR of 0.43 at 2.10 seconds into the transient.

A.1.4 MCPR Safety Limit.

For single loop operation, the NSSS ~endor found than an increase of 0.01 in the MCPR safety limit was needed to account for the increased flow measurement

• uncertainties and increased tip uncertainties ass.ociated with single pump operation.

ANF has evaluated the effects of the increased fl ow measurement uncertainties on the safety limit MCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is also applicable to ANF fuel during single* loop operation.

0.01 for single loop operation Thus, increasing the safety limit MCPR by (1.06) with ANF fuel is sufficiently conservative to al so bound the increased fl ow measurement uncertainties single loop operation.

A. l. 5 Summary It is conservative to use the reduced flow two-loop operating MCPR limit or full flow MCPR operating limit plus.01 (whichever is greatest) for single loop operations.

These limits conservatively bound all transients from SLO conditions.*

The reduced flow MCPR limit is t6 *protect against boiling transition during flow excursions to maximum two-pump flow; excursions to such high flows are not possible during single loop one-pump operation.

Thus, conservatively maintaining this 'two-loop limit assures that there is even more thermal margin under single loop conditions than under two-loop full power/full flow conditions.

A-5 ANF-87-096

-* ~

~.

.... "...:...., -~

... i,r,~,_':,

• A.2 MAPLHGR LIMITS ANF performed LOCA analyses from single loop conditions and determined an appropriate SLO MAPLHGR multiplier for ANF 8x8 and 9x9 fuels.

These ECCS analyses results are presented in "LOCA-ECCS Analysis for Dresden Units During

. Single Loop Operation with ANF Fuel,~ ANF-87-111.

A-6 ANF A.3 STABILITY Reactor operation within the limitations which assured adequate stability for the previous cycle will continue to assure adequate stability for Cycle 11.

The stability analyses

~eported in figure 4.3 of ANF-87-097 cover the operating region of Single Loop Operation; the calculated decay ratios are within the appropriate acceptable value.

A-.7 ANF-87-096 TABLE A.I SLO REACTOR AND PLANT CONDITIONS.

Reactor* Thermal Power (81.3%)

Total Recirculation Flow (58%)

Core Bypass Flow Core Inlet Enthalpy Vessel Pressures Steam Dome Lower Plenum Turbine Pressure Steam Flow Feedwater Enthalpy 2054.45 MWt.

• 56.84 Ml b/hr 5.52*Mlb/hr

. 507.8 Btu/lb 993.7 psia 1008.7 psia 958.6 psia 7.81 Mlb/hr 295.7 Btu/lb

100 CJ LlJ I-

~BO LL 0

)::>

I-I z

CX>

LlJ6Q u

a:

LlJ a..

40 20

)::>

0

z 0.0 0.5

1. 0

1. 5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 "Tl I

TIME, SEC CX>

-.....J I

0 Figure A.1 Single Loop Oper

- Pump Seizure

. 60 50 40 30 20 10 1

-1Cl u.o 2

0.5

1. 0

2.

1. 5

2. Q

2. 5 3.0 3.5

TIME, SEC Figure A.2 Single Loop Operation - Pump Seizure 4.0 4.5 5.0

)::o

z,,

I 00

-....i I

0 l.O

°'

DRESDEN UNIT 3 CYCLE 11 PLANT TRANSIENT ANALYSIS Distribution D. A. Adkisson D. J. Braun R. E. Collingham T. H. Keheley J. L. Maryott J. N.. Morgan D. F. Richey G. L. Ritter D. R. Swope C. J. Volmer R. I. Wescott J. A. White H. E. Williamson CECo/J. M. Ross (60)

Document Control (5)

ANF-87-096 Issue Date: 9/28/87