ML20153G326

Cycle 10 Plant Transient Analysis

ML20153G326
Person / Time
Site:	Dresden
Issue date:	09/12/1985
From:	Currie T, Keheley T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17199F643	List: ML20153G305 ML20153G326 ML20153G352
References
XN-NF-85-62, NUDOCS 8602280235
Download: ML20153G326 (67)

Text

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XN-NF-85-62 l

DRESDEN UNIT 3 CYCLE 10 PLANT TRANSIENT ANALYSIS SEPTEMBER 1985 RICHLAND, WA 99352 3-3 . .

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l ERON NUCLEAR COMPANY, INC.

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XN-NF-85-62 Issue Date: 9/12/85 f

DRESDEN UNIT 3 CYCLE 10 PLANT TRANSIENT ANALYSIS Prepare: b [d. (b$d T.H. Keheley, Engineer O it L a K 8

BWR Safety Analysis

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Prepare: #

-<. x # d F - 8'3 T.P. Currie, Engineer BWR Safety Analysis Concur: 'f I f u , .x in $t//j' R.E. Collihgham, Ita49er BWR Safety Analysif Concur: [ ,M .- A. 9 /idt-<-

J.N.,% rgan, Manager ~

Customer Services Engineering Approve: - M llDeN' R.B. Stout, Manager Licensing and Safety Enginaering Approve: ,0*')l{chrb>dr, Jl// . 'Wr/w G.L. Ritter, Gianager Fuel Engineering and Technical Services

XN-NF-85-62 NUCLEAR REGULATORY COMMISSION DISCLAIMER 4

IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY l

This technical report was derived through research and development programs sponsored by Exxon Nuclear Company. Inc. It is being submitted by Exxon Nuclear to the U.S. Nuclear Regulatory Commission as part of a- technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Exxon Nuclear-f abricated reload fuel or other technical services provided by Exxon Nuclear for light water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the U.S. Nuclear Regulatory Commission in its review of this report, and by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Exxon Nuclear in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any person

acting on its behalf

A. Makes aay warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the information 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 B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.

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5 i XN-NF-85-62 i

Table of Contents 1.0 INTR 000CTION................................................. ....I 2.0

SUMMARY

.......................................................... 4 3.0 TRANSIENT ANALYSES FOR THERMAL MARGIN............................ 9 3.1 DESIGN BASIS................................................ 9

3. 2 CALCU LAT IONAL M00 EL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.3 ANTICIPATED TRANSIENTS..................................... 11 i

3.3.1 Load Rejection Without Bypass....................... 11

! 3.3.2 Feedwater Flow Controll er Fail ure. . . . . . . . . . . . . . . . . . . 12 i

! 3.3.3 Loss of Feedwater Heating........................... 13 l 3.3.4 Stati stical Uncertainty Analysi s . . . . . . . . . . . . . . . . . . . . 14 T

3.4 MCPR Safety Limit.......................................... 15 4.0 MAXIMUM OVERPRESSURE ANALYSIS................................... 34-4.1 DESIGN BASIS............................................... 34 4 4.2 PRESSUR.*IATION TRANSIENTS.................................. 34 4.3 CLOSURE OF ALL MAIN STEAM ISOLATION VALVES................. 35 4

f 5.0 0FF RATED OPERATING CONDITIONS.................................. 39 5.1 AUTOMATIC FLOW CONTR0L..................................... 40 5.2 MANUAL FLOW CONTR0L........................................ 41

6.0 REFERENCES

...................................................... 49 A.1 APPENDIX A.......................................................A-1 d

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1 11 XN-NF-85-62 List of Tables 4

Table Title. Page l 2.1 DeltaCPR's................................................6 2.2 -Th'e rmal Ma rg i n Summa ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

.2 . 3 Resul ts of Pl ant Transi ent Ana,1yses . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 Design Reactor and Pl ant Condi tions . . . . . . . . . . . . . . . . . . . . . . . . 17 3.2 Significant Parameter Values Used in Analysis.. . . . . . . . . . . . .18 3.3 Control Characteristics.................................... 20-3.4 Statistical Data Used in Transient Analysis. . . . . . . . . . . . . . . . 21 3.5 Input for MCPR Safety Limit Analysi s . . . . . . . . . . . . . . . . . . . . . . . 22 1

i 5.1 Results of Off Rated Plant Transient Analysis. . . . . . . . . . . . . . 43 5.2 Automatic Flow Control..................................... 44 i j 5.3 Reduced Flow MCPR Limit for Automatic Flow Control . .. . ..... 45 5.4 Reduced Flow MCPR Limit for Manual Flow Control . . . . . . . . . . . . 46 I

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' List of Figures Figure Title Page 1.1 Power / Fl ow Operat i ng Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.1 Generator Load Rejection Without Bypass. . . . . . . . . . . . . . . . . . . . 23 4

1 3.2 Generator Load Rejection Without Bypass. . . . . . . . . . . . . . . . . . . . 24 l

3.3 Generator Load Rejection Without Bypass.................... 25 i

3.4 Feedwater Controller Failure............................... 26 4

I 3.5 Feedwater Controll er Fail ure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 t

3.6 Feedwater controller Failure............................... 28 l 3.7 Radial Power Histogram for MCPR Safety Limit Analysis...... 29 l

3.8 Safety Limi t Local Peaking for 8x8 Fuel . . . . . . . . . . . . . . . . . . . . 30 3.9 Safety Limit Local Peaking for 9x9 Fuel . . . . . . . . . . . . . . . . . . . . 31 3.10 Design Basis Local Power Distribution for ENC XN- 1 8x8 Fue1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 1

j 3.11 Design Basis local Power Distribution for -

! GE 8x8R Fue1............................................... 33

} 4.1 Containment Isolation Without Direct Scram................. 37 i

4.2 Containment Isolation Without Direct Scram................. 38

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5.1 Reduced Flow MCPR for Auto Flow Control.................... 47 5.2 Reduced Flow MCPR for Manual Flow Control.................. 48 i

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i XN-NF-85-62

1.0 INTRODUCTION

l This report describes the plant transient analyses performed by' Exxon Nuclear Co., Inc., in support of the Cycle 10 (XN-3) reload for Dresden Unit 3. This cycle is scheduled to commence operation in Summer 1986.

Cycle 10 is the third cycle during which ENC fuel will be irradiated in Dresden Unit 3. In addition to two reloads of ENC 8x8 fuel, the Cycle 10 core will contain a significant number of retrofit 8x8 assemblies fabricated by General Electric and 176 assemblies of a 9x9 lattice configuration fabricated by Exxon Nuclear. Operating limit critical power ratio values for all of these fuel types during Cycle 10 operation are established in this report. Because the ENC XN-2 SX8 fuel is mechanically identical to the ENC XN-18x8 fuel, the transient response of the two fuel types is very similar during anticipated operation; thus the two fuel types are considered to be l

the same during the transient analyses reported in this document. Similarly, since only one GE fuel type (retrofit 8x8) is to exist in this Cycle 10 core, the transient response for the General Electric fuel was explicitly modeled as retrofit 8x8 fuel.

l The analyses reported in this document were performed using the same plant transient analysis methodology (Ref. 2) as was used to establish thermal margin requirements for Cycles 8 and 9 operation of Dresden Unit 3 (Ref. 1) except for use of the following:

i o the NRC approved constant flow MCPR methodology l o the code uncertainty methodology of Ref. 8

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~2 XN-NF-85-62 I o the COTRANSA/PTSBWR updates described in Appendix A ,

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The analysis supports operation in the expanded power / flow operating map I shown in Figure 1.1. Secti_on 5.0 describes the results of the off-rated.

$ analysis performed to demonstrate that the MCPR operating limits together with the reduced flow MCPR allow operation throughout' this map.

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l 3 XN-NF-85-62 3 100% Power Line 120 "

100% Intercept Point (100/87)

(108/100)

APRM Rod Block Line 100 -

(100/100) 80 -

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l Minimum Pump Speed Line 40 I

Natural Circulation 20 .

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o 10 20 30 40 50 60 70 80 90 100 CORE FLOW (%)

Figure 1.1 Dresden 2/3 Proposed Operating Power / Flow Map

4 XN-NF-85-62 2.0

SUMMARY

The determination of thermal margin requirements for Dresden Unit 3 Cycle 10 was based on the consideration of various operational transients. The limiting transients in each general category of event are identified 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 or turbine trip event without bypass to the condenser, the loss of feedwater heating event, and the feedwater controller failure (maximum i

demand) event. For the case of Dresden Unit 3, the most limiting of these i

events was found to be the generator load rejection without bypass. Table 2.1 presents the change in critical power ratio (ACPR) at bounding conditions for the three most limiting transients.

The safety limit for Cycle 10 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 applicable to Cycle 10 operation of Dresden Unit 3 is contained in Table 2.2. These are obtained by adding the CPRs of the limiting transient in Table 2.1 to the 1.05 safety limit. The MCPR operating limit for 9X9 fuel 'is 0.04 greater than for ENC 8X8 fuel. However, for the same bundle power the operating critical power ratio (CPR) for 9X9 fuel is also about 0.04 higher than for ENC 8X8 fuel because of the larger heat transfer trea in 9X9 fuel. Thus the l

9X9 and 8X8 fuels have equivalent thermal margin to their respective MCPR limits or conversely, have the same bundle power at their respective MCPR

Y 5 XN-NF-85-62 1

operating limits.

.! Maximum system pressure for ASME overpressure evaluation has been calculated

! for the postulated closure of all main steam isolation valves (MSIVs) without activation of the MSIV position scram and without pressure relief credit for j the four electromatic relief valves. The results of this analysis as shown j in Table 2.2 indicates that the requirements of the ASME Code regarding overpressure protection will continue to be met for the Dresden Unit 3 Cycle i

) 10 core, the calculated pressures are below the 1375 psig limit.

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For informational purposes, Table 2.3 summarizes the maximum values obtained l

for other key parameters in the thermal margin and ASME overpressurization l transients.

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6 XN-NF-85-62 Table 2.1 Delta CPR's Dresden Unit 3 Cycle 10 Delta CPR Transient GE 8x8 ENC 8x8 ENC 9x9 l

l Generator .

Load Rejection Without Bypass (l) 0.23 0.24 0.28 Feedwater Flow Controller Failure 0.17 0.17 0.20 I

(Maximum Demand)(2) l l Loss of Feedwater l Heating (2) 0.20 0.20 0.20 i

(1) ACPR on statistical basis (2) ACPR on bounding basis l

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Table 2.2 1

Thermal Margin Summary Dresden Unit 3 Cycle 10 .

MCPR Operating Limit (1)

! Limiting

! Transient GE 8x8' ENC 8x8 ENC 9x9 I Generator

! Load Rejection i Without Bypass 1.28 1.29 1.33 I

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Maximum Pressurization (psta)

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Transient Steam Dome lower Plenum Steam Lines j

MSIV Closure

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- Without Position 1316 1341 1315  ;

e Scram (ASME) l 1. Based on a 1.05 safety limit. 4 i

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8 XN-NF-85-62 Table 2.3 Results of Plant Transient Analyses Dresden Unit 3 Cycle 10 Maximum Maximum Maximum Core Average Vessel Neutron Flux Heat Flux Pressure Event (% of Rated) (% of Rated) (psia)

Generator Load Rejection Without Bypass (1) 275 109.5 1273 Feedwater Flow Controller Failure 185 112.1 1191 (Maximum Demand)

Loss of Feedwater Heating 120 120.0 1064 l MSIV Closure (ASME Analysis) 494 131.7 1341 i

(1) Nominal case; all other events bounding case

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9 XN-NF-85-62 3.0 TRANSIENT ANALYSES FOR THERMAL MARGIN This text section describes the analyses which were performed to determine 1

l the minimum MCPR operating limits for Dresden Unit 3,' Lycle 10.

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i 3.1 DESIGN BASIS The plant transient analyses for Dresden Unit 3 determined that the most limiting condition for thermal margin was reactor operation at full power and flow. Reactor plant conditions for these analyses are shown in Table 3.1.

The most limiting point in the cycle is the end of full power capability, at which time the control rods are fully withdrawn from the core. The thermal margins established for the end of full power capability are conservative for l cases where control rods are partially inserted or reactor power is less than rated. 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 10 operation of Dresden Unit 3.

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10 XN NF-85-62 3.2 CALCULATIONAL MODEL The plant transient methodology described in' reference 2 and Appendix A was used for the analysis reported in this document. The COTRANSA one-dimensional core model is used to evaluate the generator load rejection and feedwater controller failure transients to model the axial power shifts associated with the system overpressurization.

The analytical models used to determine thermal margin requirements include i

PTSBWR3/COTRANSA (Ref. 2), R00EX2 (Ref. 3), and XCOBRA (Ref. 4). The

! interaction of these models to define thermal limits is described in Appendix A. Fuel pellet to cladding gap conductance values used in the analyses were based on R00EX2 calcultions for the Dresden Unit 3 Cycle 10 core configurat.an.

In accordance with ENC methodology, possible limiting transients are evaluated using a consistant 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, ENC's methodology for including code uncertainties in determining operating limits for rapid pressurization transients in BWR's (Ref. 8) is used. This methodology includes code uncertainities and uncertainties in important input variables. A conservative deterministic integral power multiplier of 110 7.

is used to account for code uncertainties when the statistical methodology is being applied.

11 XN-NF-85-62 Table 3.2 summarizes the values used for these important parameters in the an a i ., .

Table 3.3 provides the feedwater flow, recirculating coolant flow, and pressure regulation system settings used in the analysis.

3.3 ANTICIPATED TRANSIENTS Eight major categories of transients were considered generically in Reference

2. For Cycle 10 operation of Dresden Unit 3, specific events have been evaluated for thermal margin. These events are the generator load rejection transient without bypass to the condenser, and the feedwater flow co1 troller failure to maximum demand and loss of feedwater heating. In the analysis, it was assumed that a relief valve was out-of-service. For BWR/3 plants, other categories of transients are either inherently self-limiting or bounded by one of these.

i 3.3.1 Load Rejection Without Bypass This event is the most limiting of the rapid pressurization transients for Dresden Unit 3. This conclusion was verified throagh comparison with the results of the analysis of the turbine trip transient without condenser bypass for the Dresden plants.

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In the load rejection and turbine trip transients, steam flow is interrupted by an abrupt closure of either the turbine control valve in the case of the load rejection or the turbine st'op valve in the case of the turbine trip.

12 XN-NF-85-62 The resulting pressure increase causes a decrease in the vold level in the core, which in turn creates a power excursion. This excursion is mitigated in part by Doppler broadening 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, which determines the value of the scram reactivity. Other important inputs include the control rod movement parameters (scram delay and insertion speed), which determine the event characteristics following the initial mitigation of the power excursion. The bounding case resulted in delta CPR's of 0.33, 0.34 and 0.40 for the GE 8X8,, ENC 8X8, and the ENC 9X9 fuels respectively as shown in Table 5.1.

3.3.2 Feedwater Flow Controller Failure Failure of the feedwater control system could lead to a maximum increase of feedwater flow into the reactor vessel. The excessive feeddater flow increases the subcooling in the recirculating water returning to the reactor core. This reduction in average moderator temperature will result in the core power's 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 arre'sts the power increase, and the pressure

13 XN-NF-85 62 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 concurren,t, and the calculated delta-CPR is a bounding result The calculated values as shown in Table 2.1 of 0.17 for the 8x8 fuel types and 0.20 for the 9x9 fuel are adequate for protection of the fuel against boiling transition.

1 Figures 3.4, 3.5, and 3.6 illustrate the behavior of major system ' variables during the FWCF 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 assumed that the plant is operating in manual control and the feedwater ,

temperature dropped 145 degrees F over a two minute period. Void reactivity is assumed to be 25Y. Iower than the nominal calculated value, which resulted in a maximum value of the heat flux. Scram performance is assumed to be at Technical Specification limits, and control rod worth is assumed to be 20%

less than the nominal calculated value.

Previous loss of feedwater heating analysis on Dresden have shown that the delta CPR for the transient to be less limiting than the above transients, l

l Furthermore, , Reference 9. shows that the delta CPR for the transient to be the l

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I 14 XN-NF-85-62 same for 8x8 and 9x9 fuel types because it is 'a slow transient. This transient is not limiting for any fuel type in Cycle 10 of Dresden Unit 3.

The result of the loss of feedwater heating transient analysis is a delta CPR of 0.20 for all fuel types as shown in Table 2.1.

3.3.4 Statistical Uncertainty Analysis The bounding transient analysis showed the load rejection without bypass (LRW8) to be the limiting transient for Dresden Unit 3 Cycle 10, as in previous cycles. When rapid pressurization transients are limiting, ENC methodology for including code uncertainities in the determination of MCPR operating limits is applied (Ref. 8). This methodology uses a conservative deterministic multiplier of 110% on the calculated power transient and treats the uncertainities in the important input variables (scram speed and scram delay) statistically. The delta CPR used to establish the NCPR operating limits result from the use of the deterministic 110% integral power combined with 95% probability that the statistical variable delta CPR is not exceeded.

The uncertainty in the control rod drive performance parameters, scram delay and insertion speed, was determined from measured plant data. Incorporating the most recent plant data, the uncertainty in the scram delay time for Cycle 10 was determined to correspond to a mean value of 241 msec and a standard deviation of 28 msec. In the Cycle 9 analysis, a mean value of 240 msec and a standard deviation of 30 msec were used.

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15 XN-NF-85-62 Similarly, the uncertainty in the scram insertion speed for Cycle 10 was determined to correspond to a mean value of.139.57 cm/sec and a standard deviation of 3.33 cm/sec. In the Cycle 9 analysis, a mean of 136.21 cm/sec and a standard deviation of 2.74 cm/sec were used.

The uncertainties for the Cycle 9 analyses are compared with the uncertainties fer the Cycle 10 analyses in Table 3.6.

Figures 3.1, 3.2, and 3.3 illustrate the behavior of major system variables during the LRWB transient using nominal input for uncertainty variables.

3.4 MCPR SAFETY LIMIT The MCPR safety limit for Cycle 10 operation of Dresden Unit 3 was determined using the methodology described in Reference 6. No changes were made to the methodology except for code revisions required to apply the methodology to a core of mixed 8x8 and 9x9 assemblies. This methodology was used to determine the MCPR safety limit for Cycle 9 operation of Oresden Unit 2 and for Cycle 8 and 9 operation of Dresden Unit 3. As with the transient delta CPR's, the Dresden 3 Cycle 10 analysis assumed constant flow (no flow iteration) in computation of CPR as approved in Ref. 2.

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

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16 XN NF-85-62 The design basis radial power distribution used in the analysis is shown in Figure 3.7. This power distribution was found to have the greatest number of rods near limits for Cycle 10 and is conservative in :his regard relative to other expected power distributions during the cycle. lhe radial peaking for the four maximum power assemblies was increased above the calculated values to give a conservative operating NCPR of 1.40 and a significant portion of the core near operating limits. Four fuel types were represented in the Dresden 3 Cycle 10 safety limit analysis, i.e.. ENC Xh-3 9X9, ENC XN-2 8X8, ENC XN-1 8X8, and GE 8X8R.

Bounding local power distributions for each fuel type over their expected Dresden 3 Cycle 10 exposure were used. The po'ser distributions are shown in Figures 3.8 through 3.11.

The MCPR safety limit was calculated to be 1.05. Protection of this limit will assure that at least 99.9% of tha fuel rods in the core are expected to avoid boiling transition during normal operation and anticipated operational occurrences. This limit applies to the four fuel types above.

17 XN NF-85-62 Table 3.1 Design Reactor and Plant Conditions Dresden Unit 3 Reactor Thermal Power 2527 MWt Total Recirculating Flow 98.0 M1b/hr Core Channel Flow 87.7 M1b/hr l

Core Bypass Flow 10.3 M1b/hr Core inlet Enthalpy 522.9 BTU /lbm Vessel Pressures 1020 psia Steam Dome Upper Plenum 1026 psia Core 1035 psia Lower Plenum 1049 psia Turbine Pressure 964.7 psia Feedwater/ Steam Flow 9.8 M1b/hr

, Feedwater Enthalpy 320.6 BTU /1bm Recirculating Pump Flow (per pump) 17.1 M1b/hr 1

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18 XN NF-85-62 Table 3.2 f

Significant Parameter Values Used in Analysis (1)

Dresden Unit 3 High Neutron Flux Trip 3032.4 MW Control Rod Insertion Time 3.5 sec/90% inserted Control Rod Worth 20% below nominal Void Reactivity Feedback 10% above nominal (2)

Time to Deenergized Pilot Scram Solenoid Valves 283 msec (maximun)

Time To Sense Fast Turbine Control Valve Closure 80 msec Time From High Neutron Flux Trip To Control Rod Motion 290 msec Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open Turbine Control Valve Stroke Time (Total) 150 msec Fuel / Cladding Gap Conductance Core Average (Constant) 525.2 BTU /hr-ft2-F 4 (at 8.475 kW/ft)

Safety / Relief Valve Performance Settings Technical Specifications Pilot Safety / Relief Valve Capacity 166.1 lbm/sec (1080 psig)

Power Relief Valves Capacity 620.0 lbm/sec (1120 psig)

Safety Valves Capacity 1432.0 lbm/sec (1240 psig)

Pilot Operated Valve Delay / Stroke 400/100 msec Power Operated Valves Delay / Stroke 967/200 msec i

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1) LRWB transient was evaluated statistically (See 3.3.4)

2) 25*. for calculations with point kinetics model

l 19 XN-NF-85-62 l

l Table 3.2 (continued)

Significant Parameter Values Used in Analysis Dresden Unit 3 MSIV Stroke Time 3.0 see MSIV Position Trip Setpoint 90% open Condenser Bypass Valve Performance Total Capacity 1085.2 lbm/sec Delay to Opening (from demand) 100 msec Opening Time (Entire bank, max demand) 1.0 see Fraction of Energy Generated in Fuel 0.965 Vessel Water Level (above Separator Skirt)

Normal 30 in Range of Operation 20-40 in High Level Trip 48 in Maximum Feedwater Runout Flow Three pumps 4966 lbm/sec Two pumps 3310.67 lbm/sec Doppler Reactivity Coefficient (1) 0.00230 $/F-void fraction Void Reactivity Coefficient (1) -15.14 $/ void fraction Effective Delayed Neutron Fraction 0.0052 Prompt Neutron Lifetime 0.0461 msec Rectreulating Pump Trip Setpoint 1240 psig Vessel Pressure (1) Nominal value l

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i i Table 3.3 1

1-

' Control Characteristics d

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! Sensor Time Constants i Pressure 100 msec ,

Others 250 msec Feedwater Control Mode / One-Element l

i Feedwater Master Controller ,

Proportional Band ' 100% >

J 5' rep /nin j Reset '

! Feedwater 100% Mismatch ,

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Water Level Error 60 in

) 12 in eq.

! Steam Flow (not used)

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

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! Reset 8 rep / min i Speed Controller Settings -

! Proportional Band 350%

Reset 20 rep / min J i

Pressure Setpoint Adjuster Overall Gain 5 psi /% demand Time Constant 15 sec j Pressure Regulator Settings

! Lead 1.0 sec

Lag 6.0 sec

Gain 3.33 %/psid l '

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21 XN-NF-85-62 Table 3.4 Data Used in Statistical Transient Analysis Cycle 9 Cycle 10 Variable Mean Std Dev Mean Std Dev Scram Insertion Speed 136.21 2.74 139.57 3.33

, (cm/sec)

Scram Delay Time (msec) 240 30 241 28 Integral power was assumed 110 per cent of calculated value for all cases.

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1 Table 3.5 Input for MCPR Safety Limit Analysis Input Uncertainties Parameter- Standard Deviation XN-3 Correlation 0.0411 Assembly Rt. dial Peaking Factor 0.0528 Fuel Rod Local Peaking Factor 0.0246 Fuel Assembly Flow Rate 0.0280 gs Nominal Input Values Parameter Mixed Core (8x8 and 9x9)~

Core Pressure (psia) 1015 Core Power (MW) 3222 Core Inlet Enthalpy (BTU /lbm) 521.9 Total Core Flow (Mlbm/hr) ~

98.0 Feedwater Temperature (F) 345 Feedwater Flowrate (Mlbm/hr) 12.56 i- / ,

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300 1 ret umun eLo n Levet 2 HEA T FLUX 3 REC [RCULATI ON FLOW 250 h 4 3

VES 3EL STEM FLOW ret avHItM F LOW t

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0 a S I 5 0 8

0 4 -

4

, l

3.0 1 irr itw tm irtw rt wen miiJ u 2

E.,

O U

G a:

g 2.5 y a

n.

iEE.'2 o

H

) m m

F M

2.0 W i 217 ,

E

' se h

ic 1*k.0 0.5 1.0 15 2.0 2.5 3.0 3.5 4.0- 4.5 5.0 TIME. SEC l Figure 3.3 Load Rejection Without Bypass

l S EL-, h 0

4 6

3 5

3

t. W t

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, H x m

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I H

Z i H II.8 1 L t N 1 1

16 '

• 5 h

1* b 4 8 12 16 20 24 28 32 36 40 -

TINE. SEC Figure 3.6 Feedwater Controller Failure

DESIGN BRSIS RADIRL POWER DISTRIBUTION 70 -

60 - -

(n 50 -

Ed m

J cn i E La 40 - ~

En (O ru (E ,,

l ta O 30 - _

(k:

ta _ _ _ __

' cn S3 z

10 -

i $!

4 0 ,

1 0 0.2 0.1 0.6 0.8 1 1.2 1.1 1.6 1.8 R l ASSEMBLY RADIAL PEAKING ,

1 4

Figure 3.7 Design Basis Radial Power Distribution

30 XN-NF-85_62 LL  : L  : ML  : M  : H  : H  : M  : ML  : L  :

0.88 : 0.93 : 0.98 : 1.07  : 1.12 : 1.12 : 1.06 : 0.97 : 0.92 :

L  : ML  : M  : ML*  : H  : H  : ML*  : H  : ML  :

0.97 : 0.99 : 1.04 : 0.85 : 1.03 : 1.03 : 0.85 : 1.12 : 0.97 :

: H  : H  : ML'  : M  :

L  : M  : M H H

0.94 : 1.09 : 0.99 1.01 : 0.99 : 1.00 : 1.02 : 0.85 : 1.06 :

ML  : ML*  : H  : H  : H  : W : H  : H  : H  :

1.04 : 0.92 : 1.05 : 0.99 : 1.00 : 0.00 : 1.00 : 1.03 : 1.12 :

ML  : M  : H  : H  : W  : H  : H  : H  : H  :

1.04 : 1.07 : 1.05 : 1.01 : 0.00 : 1.00 : 0.99 : 1.03 : 1.12 :

L  : M  : M  : H  : H  : H  : H  : ML'  : M  :

0.95 : 1.10 : 1.00 : 1.03 : 1.01 : 0.99 : 1.01 : 0.35 : 1.07 :

L  : ML'  : ML  : M  : H  : H  : M  : M  : ML  :

0.98 : 1.01 : 0.94 : 1.00 : 1.05 : 1.05 : 0.99 : 1.04 : 0.98 :

LL  : L  : ML*  : M  : M  : ML'  : M  : ML  : L  :

0.89 : 0.95 : 1.01 : 1.10 : 1.07 : 0.92 : 1.09 : 0.99 : 0.93 :

W ..______....._____-.......__.......... .......-_............-__..._________.......

I .

0  : LL  : LL  : L  : L  : ML  : ML  : L  : L  : LL  :

E  : 0.94 : 0.89 : 0.98 : 0.95 : 1.04 : 1.04 : 0.94 : 0.97 : 0.38 :

WIOE Figure 3.8 Design Basis Local Power Distribution for ENC XN-3 9x9 Fuel t

I I

31 XN-NF-85-62

L  : ML  : ML M : M : M  : ML : ML :

0.99 : 0.99 : 0.94 : 1.06 : 1.05 : 1' . 07  : 0.94 : 0.98 :

E  : ML'  : M  : E  : H  : E*  : M  : E  :

1.01 : 0.93 : 1.04 : 1.02 : 1.01 : 0.83 : 1.04 : 0.94 :

ML  : M  : H  : H.: H  : E  : ML*  : M  :

0.99 1.07 :: 1.02 : 0.98 : 0.99 : 1.00 : 0.83 : 1 07 : '

E  : M  : H : H : W : H : H : M : ,

4

0.98 : 1,06 : 1.01 : 0.98 : 0.00 : 0.99 : 1.01 : 1.05 :

, a ML M : H : H  : H  : E  : E  : M :

0.99 : 1.07 : 1.03 : 0.9G 0.98 : 0.98 : 1.02 : 1.06 :

_ _ .. ~ __ __ __ ___ ___ _ _ _ __

1  : E  : E  : M  : E  : H  : H  : M  : E  :

1.01 :: 0.93 :: 1.05 :: 1.03 : 1.01 : 1.02 : 1.04 : 0.94 :

L E*  : E  : M  : M  : M  : ML*

• ML  :

0.99 :: 0.99 :: 0.93 :: 1.07 : 1.06 : 1.07 : 0.93 : 0.99 :

w __ __ _ _ _ -___._ _ _ . _____

I  :  :  :  :  :  :  :  :  :

D : LL  : L  : E  : E  : E  : ML  : ML  : L  :

E

0.95 :: 0.99 :: 1.01 :: 0.99 : 0.98 : 0.99 : 1.01 : 0.99 :

l WIDE J

Figure 3.9 Design Basis Local Power Distribution for ENC XN-2 8x8 Fuel

32 XN-NF-85-62

. . . . . l

t-

Mt  : ML 2 M- 'M - : *-  : ML  : ML  : 1 G.99 : D.97 : 0.95 : 1.05 .:..1 ~. 0 4 : 1.05 : 0.95 : 0.98 :

. v. . . . . . . .

. ML  :

ML. : M .: H  : H-  : ML. : M  : ML :

:.99 : 3.*A : 1 03.i 1.C3 i.--.. 1.03 : 0.88 ..: 1 44 :. C.95 :.

z. - - - - - a. - - . . . . ..

, ____....__...._............o_..........__.........__.-

i

'ML : M-  : H  :' ' H - ~H  : H ML*  : M.

r.97 : 1.06 : 1.04 : 1 01 :  : 1.02 : 0.A8 : 1.C5 :

.- 1. c l .

ML : 'M  : H  : H  : W  : H  : H ': M  :

0.97 : 1 04 : 1.c3 : 1 01 t 0.00 : 1.01 : 1.03 :. 1.04 :.

ML : M  : H  : H  : H

• H  : H  : M  :

1.97 : 1.04 : 1.33 : 1 01 : 1 01 : 1.01 : 1 33 : 1.05 :.

i l

,  : ML  : ML  : M  : H  : H  : H  : M  : ML

• l i

0.98 : 0 94 : 1.03 : 1 03 : 1.c3 : 1.34 : 1.c3 :. 0.95 :.

L  : ML. : ML  : M  : M  : M  : ML* : ML  :

0. 98 : 0.95 : 0.94 : 1 0* : 1 04 : 1.05 : 0.94 : 0.97 :

y _____....__...___..........._________..... ____.____..... .

I .

D  : LL  : L  : ML  : ML : ML  : ML : ML : L  :

E  : ' . 9 9 -: 0.?8 : 3 98 : 0.97 : 0.97 : 0.77 : 0.99 : t.a9 :

t I

W I OE Figure 3.10 Design Basis Local Power Distribution for ENC XN-1 8x8 Fuel

33 XN-NF-85-62

MH  : MH  : MH  : MH  : M  : ML  :

L  : ML

0.97 : 0.95  : 1.02  : 1.01  : 1.00  : 1.01  : 0. 96 : 0.95 :

MH*  : H  : MH  : M  :

E  : MH  : Md*  : H

0.96 : 1.02 : 0.99 : 1.06 : 0.96

1.05 : 0.99 : 0.96 :

: H  : H  : H  : MH  :

K  : MH  : H H

0.95 : 1.01 : 1.07 : 1.05 :

1.03 : 1.03 : 1.05 : 1.01 :

ML  : MH  : H  : W  : H  : H  : M*  : MH  :

0.95 : 1.01  : 1.07 : 0.00 : 1.05  : 1.03  : 0.96 : 1.00 :.

E  : Mi*  : M  : MH  : W  : H  : H  : MH  :

0.95 : 1.01  : 0.99 : 0.99 : 0.00  : 1.05  : 1.06 : 1.01 :

E  : MH  : MH  : MH  : H  : H  : MH*  : MH  :

0.% :

1.02 : 1.00 : 0.99 : 1.07

1.07

0.99 : 1.02 :

L  : ML  : MH  : W  : MH  : MH  : MH  : ML  :

0. % : 0.94 : 1.02  : 1.01  : 1.01  : 1.01  : 1. 02 : 0.95 .

W --...-----...--....--.....-..------..---.--...................-.-----..- .

I .

D  : LL  : L  : ML  : ML  : ML  : ML  : ML  : L  :

E  : 0.99 : 0. %  : 0.96 : 0.95  : 0.95  : 0.95  : 0.96 : 0.97 : .

WIDE l

Figure 3.11 Design Basis Local Power Distribution for G.E. 8x8R Fuel 1

34 XN-NF-85-62 4.0 MAXIMUM OVERPRESSURE 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, further conservatism was added by disallowing the operation of the four power-actuated relief valves as required by the ASME code. In further compliance, 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 (MSIVs).

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 clos.re of all MSIVs with failure of the MSIV position scram. The position scram, which commands reactor shutdown almost immediately upon MSIV movement, mitigates the effects of this event to the point that it does not contribute to the

35 XN-NF-85-62 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 of the MSIVs 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 events 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 a

contained within a smaller system volume than that associated with the turbine trip events. Comparative analyses have demonstrated that the containment isolation event under these conservative assumptions results in a 4

higher overpressure than either the turbine trip or the generator load

~

rejection without bypass.

j 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 limits.

i i

l

_ _ __ _ .__ ~ .. _ _ _.. .. _.._._ _. . _ _ - .

36 XN-NF-85-62 i

I i The maximum vessel pressure of 1341 psig was observed at approximately 6.5 seconds. The maximum steam line pressure of 1315 psig was observed at J

approximately 6.5 seconds. The maximum value of the sensed pressure in the steam dome was 1316 psig. The relative values of maximum pressure during the containment isolation transient indicate that the vessel and steam lines will be protected against overpressure limits defined in the ASME Code if a i

j pressure safety limit of 1345 psig in the steam dome is protected.  ;

i i

i i Figures 4.1 and 4.2 illustrate the performance of major system varia' ales ,

during the overpressurization accident. This calculation was performed with j COTRANSA.

i  !

t 3

i 3

i 1

1 1

4 i l

i I

i  ;

i I ,- . . , , _ _ _ _ .

l l

i t ,?%$

0 0

5 1 t 0 9

av' 5 _

2 t 0 5 8 L W L OW V. LO 3 t FL 2 L F W 0 N O _

t K O H L 7

F 5 U IC L- TE I XRT M e

ULS t rLU i uFCL s e n

n T [ S uACS RE va t

N L 0 6

r u

s o

tEEE e e

rHRV e 7 2 C

E l

C V

S t

\_ 0 I

S t234 3 A s 5 E

, M

^

\ M I 1

_ 5 T 4 wn s M 0 4

i e

r u

g F

C A k

@v 0 3

a, 2

r

_ t 0 s 2 4

i r

t 0 s 1 a

s z

t 0

_ 0 0 0 0 0 0 I.

1 0

0 4

0 0 0 0 1 -

6 5 4 3 2 SEg r$ G0$'

300 1 vt, - -,= == t. wen.t irm a

a. id WATE (IN1 1

x 250 7

200 a

4 150 g #

O 100 i

4 j 50 t

e a z

-0 t t i > M __ t .

c

~ _t g y 8 4%.0 1.0 2.0 3.0 1.U 5.0 6.0 7.0 8.0 9.0 10.0 k TIME, SEC Figure 4.2 MSIV Closure

l 39 XN-NF-85-62 1

5.0 ANALYSIS OF OFF-RATED CONDITIONS l

Transient analysis of a BWR requires consideration of transients at I

'off-rated' and reduced flow conditions. The MCPR full' flow operating limit is established through evaluation of anticipated transients which are i

expected to be most limiting at rated conditions. To assure that no thermal limits would be violated the generator load rejection without bypass and the f

feedwater controller falture were also evaluated at three off-rated points on i the expaned power to flow map of Fig. 1.1. The analysis indicates that all i three points are bounded by the generator load rejection without bypass at i

the 100% power 100% flow position on the operating map. The results of the analysis are summarized in Table 5.1.

i

! Analysis for pump runup events for operation at less than rated recirculation 4

pump capicity indicates the need for an augmentation of the full flow MCPR j

operating limit for lower flow conditions. This is due to the potential for large reactor power increases should an uncontrolled pump flow increase occur.

t i

i The present analysis establishes the necessary reduced flow MCPR operating limit to protect the reactor fuel against boiling transtion during i

! anticipated pump run up events from off rated core flow conditions for both l

automatic flow control and manual flow control. These-limits are shown in fi Figure 5.1 and 5.2, respectively. 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 control made and full flow cycle i

. ~ \

1 i

40 XN-NF-85 specific MCPR operating limit.

5.1 AUTOMATIC FLOW CONTROL If the reactor is operated in the ~ automatic flow control mode (AFC),

variations in core power should not result in critical power ratios less than the established MCPR operating limit for rated conditions. If the rated condition MCPR limit is observed in a reduced flow condition, a subsequent increase in power to full power along the AFC control line may result in 3

inadvertent degradation of fuel critical power ratios to below this referer.ce(full power) MCPR operating limit. The probability of boiling transition conditions occurring during a subsequent anticipated event may increase beyond acceptable levels if this were the case.

Exxon Nuclear Company has determined the required reduced flow MCPR operating limit for off-rated conditions to prevent the MCPR from degrading below the >

Cycle 10 MCPR. (full flow) operating limits during AFC operation. This was 3 determined by evaluating the MCPR for a given reactor power distribution at varying total reactor power and flow conditions. The variations in total i

core power and flow were assumed to follow the expected relationship (Table 5.2) for automatic flow control operation (100% rod line). The power distribution chosen was such that MCPR equaled the referenced MCPR operating limit at rated conditions of power and flow. The expected variation of core pressure and inlet coolant subcooling with reactor power level was also considered. .

, ~ ,- ,

41 XN-NF-85-62' The reduced flow MCPR's were then calculated by XCOBRA (Ref. 6) along .the 100% rod line. The reduced flow MCPR limit for AFC is presented in Figure 5.1 and Table 5.4.

5.2 MANUAL FLOW CONTROL This section dicusses pump excursions when the plant is not in automatic flow control operation mode.i.e., manual flow control. Based on the results obtained from the Dresden Unit 3 analysis which showed two pump excursions were the limiting pump run-up event, only two pump excursions are evaluated for Dresden Unit 3 Cycle 10. These results indicate that MCPR would decrease below the safety limit if the full flow reference MCPR was observed.at' initial conditions. Thus, an augmented MCPR is needed for partial flow operation to protect the two pump excursion event.

The evaluation of the two recirculation pump flow excursion for Dresden Unit 3 showed that establishment of MCPR limits for this event which prevents boiling transition will also bound single pump failures. The analysis of the two pump flow excursion indicates that the limiting event scenario is a gradual quasi-steady run up due to the inlet enthalpy lag associated with a more rapid run-up.

1-a i

l 42 XN-NF-85-62 L-l' The analysis conservatively assumed the reactor reaches 120% rated power at 110% rated fl ow. This power to flow relationship bounds that calculated by .

XTGBWR for the constant Xenon assumption. The reduced flow MCPR calculations i

j- were performed assuming the event was initiated from the APRM Rod Block Line 1 as well as the 100% flow control line. ..The results show that pump run-up l

' events initiated from the 100% flow control line are bounding.

i i

The results of the two pump run-up analyses for manual flow control are fl presented in Figure 5.2 and Table 5.4. The cycle specific MCPR-limit for l

l

)

i Dresden Unit 3 shall be the maximum of the reduced flow MCPR operating limit i >

and the full flow MCPR operating limit. ,i i ,

r 1

f t

1 i

1 i

I 5

43 XN-NF-85-62 Table 5.1 Bounding Delta CPR's at Off Rated Conditions DRESDEN UNIT 3 CYCLE 10 Boundirg Delta CPR Transient GE 8x8 ENC 8x8 ENC 9x9 100% POWER 100% FLOW Generator Load Rejection 0.33 0.34 0.40 Without Bypass Feedwater Flow Controller Failure (Maximum Demand) 0.17 0.17 0.20 100% POWER 87% FLOW Generator Load Rejection Without Bypass 0.30 0.31 0.37 Feedwater Flow Controller Failure (Maximum Demand) 0.13 0.13 0.15 85% POWER 61% FLOW Generator Load Rejection With Bypass 0.31 0.31 0.38

Feedwater Flow

! Controller Failure (Maximum Demand) 0.18 0.19 0.24 l ,

! 67% POWER 39% FLOW l Generator Load Rejection Without Bypass 0.12 0.12 0.14 Feedwater Flow Controller Failure (Maximum Demand) 0.08 0.08 0.09 l

l

44 XN-NF-85-62 Table 5.2 Automatic Flow Control Recirculating Flow Power

(% Rated) (% Rated) 100 100 90 94 80 88 70 81 60 74 i 50 67 40 58

45 XN-NF-85-62 Table 5.3 Reduced Flow MCPR Limits for Automatic Flow Control Recirculating Flow MCPR Limit

% Rated) 8x8 9x9 100 1.29 1.33 90 1.32 1.35 80 1.35 1.38 70 1.40 1.42 60 1.45 1,48 50 1.52 1.54 40 1.65 1.67 e

l 46 XN-NF-85-62 Table 5.4 Reduced Flow MCPR Limits for Manual Flow Control Recirculating Flow MCPR Limit

% Rated) 8x8 9x9 100 1.10 1.09 90 1.15 1.14 80 1.21 1.19 70 1.27 1.25 60 1.35 1.32 50 1.44 1.42 40 1,58 1.53

1.8 . -

9 1,7 . .

1 9x9 t- 1.6 , ,,

E, e

5 8x8 E 1.5 . . i 8 e E

o r 1,4 , ,

I 4

1, . 3 .

5 p.2 . . . . . . . . k.

20 3D 40 SD 60 70 80 90 100 $

b TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR) 1 Figure 5.1 Reduced Flow MCPR for Auto Flow Control

{

i d

1. 6, i 8x8 1.5, ,

9x9 x

C 1.4 l e 5

W t

8 1. 3,,

E o

K '

e- y_

m

1.2 ,

9 i

1.1, ,

=

. ~ ,_ .

./

_ 2

~ n 1.0

;

: 'J:  : e  : &

20 30 40 50 60 .. 70 -80 90 100 T s j g TOTAL CORE RECIRCULATING FLOW (% RATED, 98 MLB/HR)

Figure 5.2 Reduced Flow MCPR for Manual Flow Control

~. - . . .,

49 XN-NF-85-62

6.0 REFERENCES

1. R.H. Kelley, " Plant Transient Analysis for Dresden Unit 3, Cycle 10,"

, XN-NF-83-58, Exxon Nuclear Co, Inc., Richland, WA 99352 (August 1983).

4'

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

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

4. T.L. Krysinsk'i and J.C. Chandler, " Exxon Nuclear Methodology for Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary Description," XN-NF-80-19(P), Volume 3, Revision 1, Exxon Nuclear Co.,

Inc., Richland, WA 99352 (April 1981).

5. T.L. Krysinski et al., " Exxon Nuclear Methodology for Boiling Water Reactors; Neutronics Methods for Design and Analysis," XN-NF-80-19(A),

Volume 1, Exxon Nuclear Co., Inc., Richland, WA 99352 (May 1980).

J 6. T.W. Patten, " Exxon Nuclear Critical Power Methodology for Boiling Water l Reactors," XN-NF-524(P), Revision 0, Exxon Nuclear Co., Inc., Richland, WA 99352 (November 1979).

l 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 i

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).

t

A-1 XN-NF-85-62

-APPENDIX A

. MODIFICATIONS TO COTRANSA/PTSBWR3 1.0 GENERIC CODE UPDATES COTRANSA originated with the coupling of ~ a pl ant transient simulator code, PTSBWR3, and a one dimensional, . coupled neutronic hydraulic code, COTRAN. Subsequent to the licensing of Dresden-3 Cycle 9 the following modifications have been introduced into ENC's BWR plant transient model:

o latest version of COTRAN replaced the original COTRAN o centrol system module introduced to coding o codes COTRAN, COTRANSA, PTSBWR3, and CONTROL all reside in the same program library The latest vercion of COTRAN (JUL83) repl aced the original COTRAN because the numerical convergence features had been upgraded to increase code execution efficiency. In addition the code core- outflow is now deterministically calculated insited of assumed equal to the 2 inlet flow. A Control System Module h 5 rati ,yt i the original -control i

system model so that all operations are handled through the input stream and may be tailored to the user's specific needs. This allows the user to simulate a reactor control system different than a pre-defined model . Having COTRAN, COTRANSA, PTSBWR3, and CONTROL in the same program library permits stand alone or grouped execution of each of the cod,es. It also allows individual or grouped modifications as required for application purposes.

4

-wm, , , , , - , , , , , , , - , < - - - - - - - ~nw, -,,v,~,,-,, , w w enw

A-2 XN-NF-85-62 2.0 PLANT SPECIFIC CHANGES TO COTRANSA/PTSBWR3 2.1

SUMMARY

The ENC plant transient analysis codes incorporata plant specific features through input and coding. changes. This section discusses the required Dresden Unit 3 plant specific changes and control system input to the COTRANSA/PTSBWR3 plant simulator codes. These modifications were incorporated for exclusive use to Dresden Units 2 and 3. Unless otherwise stated, modifications that were made to the PTSBWR3 portion of the code were paralled in the COTRANSA portion.

2.2 FEEDWATER CONTROLLER Because of the modifications made to COTRANSA the feedwater control system is incorporated into the code through the input stream.

Figure A-1 gives a schematic of the control system used.

The feedwater control system maintains a pre-established level in the reactor vessel during nomal plant operation by varying the speed of the steam turbine driven feed p eps. Steam flow and feed flow are compared and an error signal is sent to the mismatch gain amplifier.

The sensed reactor water level is capared to the level setpoint, this error signal is summed with the mismatch gain amp 1fier signal to provide the input signal to the flow controller. The flow controller provides the input to the function generator after going through an output limiter and a lead / lag compensator. The function generator signal is then sent to the turbine feed pumps.

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l A-3 XN-NF-85-62 2.3 RECIRCULATION FLOW CONTROL SYSTEM a

To determine if the manual or automatic mode of recircul ation is i

more limiting, the loss of feedwater heating transient is analyzed in both methods of control. The recirculation flow control system is codelled as described in Reference 2. A diagram of the system is shown in figure A-2.

2.4 PRESSURE REGULATOR CONTROL SYSTEM As is discussed in Appendix A, the pressure regulator control systen is entered into COTRANSA/PTSBWR3 as input data... The model as it -

is input is shown in Figure A-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 with a pressure setpoint. This error signal is conditioned by the lead / lag characteristics of the control valve and produces a steam flow based on the pressure setpoint.

3.0 COTRANSA HOT CHANNEL MODEL i

3.1

SUMMARY

1 The original COTRANSA hot channel model was used to give a figure of merit del ta. CPR used to determine the limiting transient. The limiting delta CPR was then determined by the user using a cumbersome

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XCOBRA-RODEX2-HUXY manual iteration. This involved a time consuming

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A-4 XN-NF-85-62 i

calculation where the user manipulated a considerable amount of data between codes. Furthemore, it also resulted in a transient condition being analyzed using steady state approximations. The COTRANSA hot l channel model has been modified to automate the' delta CPR calculation and to give a transient delta CPR. Each fuel type is model ed, and a delta CPR specific to that fuel type is detemined. XCOBRA and R00EX2 are used to determine the input for each hot channnel . COTRANSA then calculates the delta CPR for each time step. The largest delta CPR is then reported.

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3.2 MODIFICATIONS TO THE HOT CHANNEL I

3.2.1.F1ow Responce Surface 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 flow to the hot channel is detemined using XCOBRA,' ENC's

} approved subchannel , code for BWR's. A steady state response surface for the hot assemblies' flow rates are detemined for four key j variables:

) o Relative assembly themal power l

) o Core average themal power o Core average active flow i

o Core pressure A quadratic equation is then detemined for each hot assembly flow.

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A-5 XN-NF-85-62 i

3.2.2 Fuel Tsuperature Model .

The fuel temperature model for the hot rod is as described in the approved HUXY (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 average fuel tenperature is then determined. This gap 3

conductance is then used in the hot channel model.

3.2.3 Critical Power Ratio Calculations 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 wi th the core average solutions since the impact of the hot assembly is so snali. 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 o Bulk fluid temperature o Clad to fluid heat transfer coefficient o Heat flux o Axial power distribution o Enthalpy rise

A-6 XN-NF-85-62 The purpose of the calculation is to determine the maximum allowable assembly power which will not exceed critical heat flux conditions during the transient.

The initial power used in the calculation is only 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 assuned as an initial condition. The flow to the limiting assembly is determined from the response surface and the enthalpy rise is adjusted to be consistent with the new 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.

4.0 VERIFICATION OF THE HOT CHANNEL Two different checks were made to insure the adequacy of the COTRANSA hot channel model's del ta CPR calculation. The standard XCOBRA - R00EX2 - HUXY iteration was perfonned, and steady state conditions were input into the hot ' channel . The XCOBRA - RODEX2 - HUXY iteration resulted in del ta CPR's that were more conservative than those for the hot channel model . This is what was expected because of the steady state nature of the method. When quasi-steady state conditons were . forced into the hot channel , the resul ts of the comparision were favorable and as expected.

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XN-NF-85-62

-Issue Date: 9/12/85-

'i DRESDEN UNIT 3 CYCLE 10 PLANT. TRANSIENT ANALYSIS Distribution J.C. Chandler i R.E. Collingham T.P. Currie S.E. Jensen T.H. Keheley T.L. Krysinski -

J.N. Morgan L.C. O'Malley.

G.L. Ritter R.H. Schutt R.B. Stout D.R. Swope 1

- R.I. Wescott H.E. Williamson CECO / J.M. Ross (60)

Document Control (5) ,

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