nonproprietary version of revision 1 to 'plant transient

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XN-NF-82-84(NP) .

Revision 1 I

Issue Date: 12/07/82

PLANT TRANSIENT ANALYSIS FOR

DRESDEN UNIT 2 CYCLE 9

Prepared by: 2/2/I2 )'

R. H. Kelley / '

Plant Transient AnalysisU

Concur: /7 />/# t-G. C. Cooke, ManagerPlant Transient Analysis

10 9Approve: IN d

.

a ch. 6R. B. Stout, ManagerLicensing & Safety Engineering

#./~~ -,,

'- '~

' ' ' ~Approve-G. A. Sofer,' ManagerFuel Engineering & Technical Services

gf

E(ON NUCLEAR COMPANY,Inc.

Sah2pagpagagg;P

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NUCLEAR REGULATORY COMMISSION DISCLAIMER

_lAAPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT

PLEASE READ CAREFULLY

This technical report was derived through research and developmentprograms sponsored by Exxon Nuclear Company, Inc, it is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC whichutilire Exxon Nucleer fabricated reload fuel or other technical nevicesprowded by Exxon Nuclear for licht water power reactors and it is trueand correct to the best of Exxon Nucteer's knowledge, infornution,and belief. The information contained herein may be used by the USNRC

_ in its review of this rep)rt, and by licensees or applicants before theUSNRC which are customers of Exxon Nuclear in their demonstrationof complisnee with the USNRC's regulations.

Without derogating from the foregoirg neither Exxon Nuclear norany person acting nn its behalf:

A. Makes any warranty, express or implied, with respect tothe act.uracy, completeness, or usefulness of the infor-motion contained in this document, or that the use of

any informatiort apparatus, method, or promes disclosedin this document will not infringe privately owned rights;or

B. Assumes any liabilities with respect to the use of, or fordarrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN- NF- F00,766

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XN-NF-82-84(NP)i Revision 1li

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TABLE OF CONTENTS

SECTION PAGE

1.0 INTRODUCTION ............................................. I

2.0 SUMMARY 2.................................................

3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 5...................

4.0 MAXIMUM OVERPRESSURIZATION ............................... 24

5.0 REFERENCES ............................-.................. 20

Appendix A .................................................... A-1

Appectix B .................................................... B-1

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LIST OF TABLES

TABLE PAGE

2.1 Thermal Margin ............................................ 3

2.2 Resul ts of Pl ant Transient An alyses . . . . . . . . . . . . . . . . . . . . . . . 4

3.1 Design Reactor and Plant Conditions (Dresden 2) ........... 10

3.2 Significant Parameter Values Used ......................... 11

3.3 Control Characteristics ................................... 13

A-1 Dresden-2 Cycle 9 Safety LimitFuel-Related Uncertainties ................................ A-3

'

A-2 Dresden-2 Cycle 9 Safety Limit '

Nominal Input Parameters .................................. A-4

B-1 ACPR Experimental Design .................................. B-2

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LIST OF FIGURES.

FIGURE PAGE

143.1 Scram Reactivity ........................................

3.2 Axial Power Distribution 15...............................

3.3 Generator Load Rejection w/o Bypass ...................... 16(fxpected Power and Flows)

3.4 Generator Load Rejection w/o Bypass ...................... 17(E.pected Vessel Pressure and Level)

3.5 Generator Load Rejection w/o Bypass ...................... 18(Expected CPR for a Typical Fuel Assembly)

3.6 Increase in Feedwater Flow (Power and Flows) ............. 19

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3.7 Increase in Feedwater .................................... 20

3.8 Increase in Feedwater Flow (Typical CPR) ................. 21

3.9 Loss of Feedwater Heating (Power and Flows) .............. 22

3.10 Loss of Feedwater Heating(Vessel Pressure and Level) .............................. 23

4.1 MSIV Closure without Direct Scram(Power and Flows) ........................................ 26

4.2 MSIV Closure without Direct Scram27(Vessel Pressure and Level) ..............................

A-1 Dresden-2 Cycle 9 Safety Limit RadialPower Histogram .......................................... A-5

A-2 Dresden-2 Cycle 9 Safety Limit LocalPeaking .................................................. A-6

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1.0 INTRODUCTION

This report preser;ts the results of Exxon Nuclear Company's (ENC)

evaluation of core-wide transient events for Dresden Station Unit 2 during

Cycle 9 operation. Specifically, the evaluation determines the necessary

thermal raargin limits to protect against the occurrence of boiling

transition during the most limiting anticipated transient. Also, the

evaluation demonstrates that vessel integrity will be protected during the

most limiting pressurization event. The results are also incorporated in

Reference 2.

This analysis was performed with the same methodology (l) used to

establish thermal margin requirements for Dresden Unit 3 Cycle 8. The1

limiting expected transient, load rejection without condenser bypass, and

maximum pressurization event, closure of all main steam isolation valves,

were determined to be the same for Dresden Unit 2 as previously determined

for Dresden Unit 3(6),

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2.0 SUMMARY

The determination of the Minimum Critical Power Ratio (MCPR) for,

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Dresden Unit 2 Cycle 9 was based upon the consideration of various possible

operational transients (l). A MCPR of 1.31 or greater for all 8x8 fuel

types during Cycle 9 adequately limits the occurrence of boiling transi-

tion during an end of cycle full load rejection without condenser bypass,

as well as other less limiting anticipated operational transients. This

assumes compliance with other related restrictions specified by the

Dresden Unit 2 Operating License and associated Technical Specifications.

The PCPR operating limits required for the more potentially limiting

events are shown in Table 2.1.

The maximum system pressure has been calculated for the containment

isolation event, which is a rapid closure of all main steam inlation

valves, with an adverse scenario as specified by the ASME Pressure Vessel

Code. The safety valves of Dresden Unit 2 have sufficient flow capacity

and opening rates to prevent pressure from reaching the established

transient safety limit of 1375 psig, which is 110% of design pressure. The

maximum system pressures predicted during the event are shown in Table 2.1.

A sumary of results of the transient analyses is shown inTable 2.2.

This Table shows the relative maximum fuel power levels, core ' average heat

fluxes, and maximum vessel pressures attained during the more < limiting

transient events.

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3 XN-NF-82-84 (NP)Revision 1

Table 2.1 Thermal Margin Sumary

Transient ACPR/MCPR

8x8(XN-1) 8x8R(GE) 8x8(GE)

Generator .26/1.31(1) .26/1.31(1) .26/1.31(1)Load Rejection(w/o bypass)

Increase in .21 .21 .21Feedwater Flow

Loss of .16 .16 .16Feedwater Heating

MaximumPressure(psig)*

Transient Vessel Dome Vessel Lower Plenum Steam Lines

MSIV C1osure 1324.9 1349.3 1322.1

* Limit allowed is 1375 psig

_____________________________,

(1) See Section 3.2.1 for basis of these values ,

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Table 2.2 Results of Plant Transient Analyses

Event Maximum Maximum MaximumNeutron Flux Core Average Vessel(% Rated) Heat Flux Pressure

(% Rated) (psig)

g d g tion (1) 350.3 114.5 1281.1

Increase in 298.4 116.9 1207.2'

Feedwater Flow

. *Loss of Feed- 120.1 (120.0 1039.9water Heating

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I MSIV Closure 483.3 133.0 1349.3'

w/ flux scram

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i (1) Nominal case, all other events are bounding caseI:

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: 82~ &.;

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3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN

3.1 DESIGN BASIS

The plant transient analysis determined that the most thermal

margin limiting condition was operation at full reactor power. Reactor and

plant conditions for this analysis are shown in Table 3.1. The most

limiting point in cycle was end of full power capability when control rods

are fully withdrawn from the core. The thermal margin limit established

for end of full power conditions is conservative for cases where control

rods are partially inserted or reactor power is less. Following

requirements established in the Plant Operating License and associated

Technical Specifications, observance of the MCPR opernting limit of 1.31

or greater for all 8x8 fuel types protects against boiling transition

during all anticipated transients at the Dresden Unit 2 for Cycle 9.

The calculational models used to determine thermal margin

include ENC's plant transient (1), fuel performance (4), and core thermal-

hydraulic (5) codes as described in previous documentation (l). Fuel pellet

to clad gap conductances used in the analyses are based on previously

submitted analyses (6). All calculational models have been benchmarked

against appropriate measurement data, but the current evaluations are

intentionally designed to provide a thermal margin which accobnts for the

random variability and uncertainty of critical parameters. For the

limiting generator load rejection without bypass event, the variability of

four critical parameters was statistically convoluted so that the calcul-

ated thermal margin bounds 95% of the possible outcomes. Table 3.2

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sumarizes the values used for important parameters. Table 3.3 provides

the feedwater flow, recirculating coolant flow, and pressure regulation

system settings used in the evaluation.

3.2 ANTICIPATED TRANSIENTS

ENC considered eight categories of potential transient occur-

rences for Jet Pump BWR's in XN-NF-79-71(1). Three of these transients,

have been evaluated here to determine the thermal margin for Cycle 9 at

Dresden Unit 2. These transients are:

generator load rejection w/o bypass.

increase in feedwater flow.

loss of feedwater heating.;

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Other plant transient events are inherently non-limiting or

clearly bounded by one of the above.

3.2.1 Generator Load Rejection without Condenser Bypass

This event is the most limiting of the class of transients

characterized by rapid vessel pressurization. The turbine / generator

control system causes a fast closure of the turbine control valves.

Closure of these valves causes the reactor system to be pressurized while

the reactor protection system scrams the reactor in response to the sensing

of the fast closure of the control valves. Condenser bypass ' flow, which

can mitigate the pressurization effect, is not allowed. The excursion of

core pover due to void collapse (by pressurization) is terminated by

| reactor scram since other mechanisms of power shutdown (Doppler feedback,|

j pressure relief, etc.) are only partly successful. Figures 3.3, 3.4 and

3.5 depict the time variance of critical reactor and plar.t parameters

j during a load rejection event with expected void reactivity feedback and

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7 XN-NF-82-84 (NP )Revision 1

normal scram performance. ENC evaluated this event to determine a ACPR

which would not be exceeded in 95% of the possible outcomes of the event

whenfourva,ripleswereconsidered: _

._ -.

The standard deviations of the first two variables were~

The standard deviations of the latter two'

variables were based upon plant test data:__ _.

-_.

The experimental design for the statistical analysis is given in Appen-

dix B. The calculated results of the statistical evaluation were:

mean ACPR .232

standard deviation .014'

95% ACPR .260

3.2.2 Increase in Feedwater Flow' Failure of the feedwater control system is postulated to

lead to a maximum increase of feedwater flow into the vessel. As the

excessive feedwater flow subcools the recirculating water returning to the

reactor core, the core power will rise and attain a new equilibrium if no

other action is taken. Eventually, the inventory of water in the downcomer

* Brackets identify ENC proprietary information.

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Revision 1 (NP)XN-NF-82-848

will rise until the high vessel water trip setting is exceeded. To protect

against spillover of subcooled water to the turbine, the turbine trips,

w ) cesultant closure of the turbine stop valves. The power increase is

terminated by scram, and pressure relief is obtained from the bypass valves

opening. The present evaluation of this event assumed that all the

conservative conditions of Table 3.2 were concurrent; no statistical

evaluation was considered, and the ACPR calculated represents a bounding

result. Though small differences exist between G.E. and ENC fuel, the

highest ACPR of 0.21 reported is adequate to protect all fuel types against

boiling transition. Figures 3.6, 3.7 and 3.8 display critical variables

for this event for the critical 4 seconds following the turbine trip.

3.2.3 Loss of Feedwater Heating

Tte loss of feedwater heating leads to a gradual increase

in tha subcooling of the water in the reactor lower plenum. Reactor power

slowly rises to the overpower trip point (120% of rated power). The

gradual power change allows fuel thermal response to maintain pace with the

increase in neutron flux. For this analysis, it was assumed that the

initial feedwater temperature dropped 1450F linearly over a two minute

period. The magnitude of the void reactivity feedback was ass.umed to be

25% lower than expected, so that the power response to subcooling was

gradual, maximizing the thermal heat flux. Scram performance was assumed

at its Technical Specification limit with scram worth 20% below expected.

Reactor neutron flux reached 120.1% of rated and clad surface heat flux

increased nearly as much. Calculation of thermal margin assumed that

bundle power increased by 20% which predicted a ACPR of 0.16 for each fuel

| typat. Figures 3.9 and 3.10 depict the transient progression.1

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3.3 CALCULATIONAL MODEL

The plant transient model used to evaluate the load rejection and

feedwater increase event was ENC's advanced code, COTRANSA(1). This one-

dimensional neutronics model predicted reactor power shifts toward the core

middle and top as pressurization occurred. This was accounted for explicitly

in determining thermal margin changes in the transient. The loss of feedwater

heating event was evaluated with the PTSBWR3(1) code since rapid pres-

surization and void collapse do not occur in this event.

3.4 SAFETY LIMIT

The safety limit is the minimum value of the critical power rat o

(CPR) at which the fuel could be operated, where the expected number of rods

in boiling transition would not exceed 0.1% of the heated rods in the core.

Thus, the safety limit is the minimum critical power ratio (MCPR) which would

be permitted to occur during the limiting anticipated operational occurrence

as previously calculated. The MCPR operating limit is derived by adding the

change in critical power ratio (eCPR) of the limiting anticipated operational

occurrer.ce to the safety limit.

The safety limit for Dresden Unit 2 Cycle 9 was determined by the

methodology presented in Reference 3, and used to license Dresden-3, to have

i the following value- |

|Dresden Unit 2 Cycle 9 MCPR Safety Limit = 1.05. I

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The input parameter values and uncertainties used to establish the safety

limit are presented in Appendix A.

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Table 3.1 Design Reactor and Plant' Conditions (Dresden 2)1

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Reactor Thermal Power (Mwt) 2527.0

TotalRecirculatingFlow(M1b/hr) 98.0

Core Channel Flow (Mlb/hr) 87.4

Core Bypass Flow (Mlb/tr) 10.6 ;' '

Core Inlet Enthalpy (SIU/lbm),l 522.9 ,

Vessel Pressures (psia) i

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Dome 1020.0

Upper Plenum 1026.0

Core 1035.0:

Lower Plenum 1049.0

Turbine Pressure (psia) 964.79

Feedwater/ Steam Flow (Mlb/hr) 9.8 g

yfeedwater Enthalpy (BTU /lbm) 320.6

Recirculating Pump Flow (Mlb/hr) 17.1 (1) -

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=

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E

II) Per pump

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F Table 3.2 Significant Parameter Values Used (1)_

[ t High Neutron Flux Trip 3032.4 MW

Control Rod Insertion Time 3.5 sec/90% inserted,

=

Control Rod Worth 20% below nominali

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- Void Reactivity Feedback 10% above nominal (2)'

Time to Deenergized Pilot Scram 283 msec_ ,

' ~

_ Solenoid Valves'

'

s Tfce to Sense Fast Turbine 80 msec,

Control Valve Closure-

Time from High Neutron Flux 290 msec

Trip to Control Rod Motion'

aYi;; Turbine Stop Valve Stroke 100 msec

iF= ; Turbine Stop Valve Position Trip 90% open

Turbine Control Valve Stroke 150 msec

$ (Total)L

p Fuel / Clad Gap Conductance

k Corc Average (Constant) 893 BTU /hr-ft OF2

k Limiting Assembly 1430 BTU /hr-ft .or2

[ (variable *) (at8.475kw/ft)2 .* Safety / Relief Valve Performance

Settings Technical Specifications

rn

(1) Generator load rejection w/o bypass event was evaluatedstatistically (see Section 3.2.1)

(2) 25% for calculations with point kinetics model=

Varies slightly with power and fuel type*

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|ITable 3.2 Signiffcant Parameter Values Used (cont.)

Safety /ReliefValvePerformance(cont.)Pilot Safety / Relief Valve Capacity 166.1 lbm/sec (at 1080 psig)Power Relief Valves Capacity 620.0 lbm/sec (at 1120 psig)Safety Valves Capacity 1432.0 lbm/sec (at 1240 psig)Pilot Operated Valve Delay / Stroke 0.4/0.1 secPower Operated Valves Delay / Stroke 0.65/0.2 sec

MSIV Stroke Time 3.0 sec jMSIV Position Trip Setpoint 90% open '

Condenser Bypass Valve Performance |'

Total Capacity 1085.2 lbm/sec+ Delay to Opening (from demand) 0.1 secOpening Time (Entire Bank with 1.0 sec

(Maximum Demand)

% Energy Generated in Fuel 96.5%

Vessel Water Level (above Separator Skirt)Normal 30 inchesRange of Operation +10 inches

High Level Trip 42 inchesMaximum Feedwater Runout Flow (3 pumps) 4966 lbm/secMaximum Feedwater Runout Flow (2 pumps) 3310.67 lbm/secDoppler Reactivity Coefficient (nominal) -0.002325/0F/ void fractionVoid Reactivity Coefficient (nominal) -16.40$/ void fraction

i Scram Reactivity Worth Figure 3.1 -Axial Power Distribution Figure 3.2Delayed Neutron Fraction .0051

Prompt Neutron Lifetime 4.93 x 10-5 secRecirculating Pump Trip Setpoint 1240 psig (vessel pressure)

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Table 3.3 Control Characteristics

Sensor Time Constants

Pressure 0.1 sec

Others 0.25 sec

Feedwater Control Mode 1-element

Feedwater Master Controller

Proportional Band 100%

Reset 5 repeats / min

Feedwater 100% Mismatch

Water Level Error 60 inches

Steam Flow (not used) 12 in equivalent

Flow Control Mode Master Manual

Master Flow Control Settings



Speed Controller Settings



Pressure Setpoint Adjustor

Overall Gain 5 psi /% demand

Time Constant 15 sec

Pressure Regulator Settings'

Lead 1.0 sec

Lag 6.0 sec

Gain 30 psid/100% demand

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FRACTION OF CONTROL ROD INSERTION 1

Figure 3.1 Scram Reactivity

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Fraction of Active Fuel (From Bottom)

Figure 3.2 Axial Power Distribution

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Figure 3.5 Generator Load Rejection w/o Bypass (Typical CPR)______ ____ _-- - _

_ ______________. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

300 1 retuinuri et.ug

E. HEAT FLUX3. RECTRCULATION R.0W4. YESSEL STEAM FLOW

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Figure 3.6 Increase in Feedwater Flow (Power and Flows from 20-24 Seconds)

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4.0 MAXIMUM OVERPRESSURIZATION

4.1 DESIGN BASIS

The reactor conditions used in the evaluation of the maximum

pressurization event are those shown in Table 3.1. In addition to the

conservative assumptions shown in Table 3.2, ENC assumed that the four power

actuated relief valves were not available to vent steam as the ASME Pressure

Vessel Code does not allow credit for power operated relief valves. Also, the

most critical active component (scram on MSIV closure) was failed during the.

transient.

4.2 PRESSURIZATION TRANSIENTS

ENC has evaluated several pressurization events, and has determined

that closure of all main steam isolation valves without direct scram is most

limiting for maximum vessel pressure. Though the closure rate of the MSIVs is

substantially slower than turbine stop or control valves, the compressibility

of the fluid in the steam lines causes the severity of the compression wave of

the slower closure to be nearly as great as the faster turbine stop or control

valves closures. Essentially, the rate and magnitude of steam velocity

reduction is concentrated toward the end of valve stroke, generating a

substantial compression wave. Once the containment is isolated, theI

subsequent core power production must be absorbed in a smaller volume than if

turbine isolation occurred. Calculations have determined that the overall

result is to cause containment isolation to be more limiting than turbine

isolation.

4.3 CLOSURE OF ALL MAIN STEAM ISOLATION VALVES

This calculation assumed all four steam lines were isolated at the

containment boundary within 3 seconds. Due to the valve characteristics and

|

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-. - s, , - , - , , , * - --- - , - - - . , p--_ p- ,,---. ,


steam compressibility, the vessel pressure response is not noted until about

3 seconds after beginning of valve stroke. Since scram performance was

degraded to its Technical Specification limit for this analysis, effective

power shutdown is delayed until after 5 seconds. Due to limitations in steam

venting capacity, (i.e. power operated relief valves fsilures), significant

pressure relief is not realized until after 5 seconds, preventing that

mechanism from assisting in power shutdown. Thus, substantial thermal power

production enhances the pressurization. Pressures reach the recirculating

pump trip setpoint (1240 psig) before the pressurization has been reversed by

the lifting of the safety valves. Loss of coolant flow leads to enhanced steam

production as less subcooled water is available to absorb core thermal power.

The maximum pressure calculated in the steam lines was 1337 psia occurring

near the vessel at about 6.75 seconds. The maximum vessel pressure was 1364

psia occurring in the lower plenum at about 6.5 seconds. Figures 4.1 and 4.2

illustrate the progression of the transient.

The calculation was performed with ENC's advanced plant simulator

code, COTRANSA, which includes a one-dimensional neutronics model.

,

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Figure 4.2 MSIV Closure without Direct Scram (Vessel Pressure and Level)

|

28 . XN-NF-82-84 (NP)Revision 1

5.0 REFERENCES

(1) " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors",XN-NF-79-71(P), Revision 2, Exxon Nuclear Company Inc., Richland,Washington 99352, November 1981.

|

!

(2) "Dresden Unit 2 Cycle 9 Reload Analysis", XN-NF-82-77(P), Revision 1,,

| Exxon Nuclear Company Inc., Richland, Washington 99352, December'

1982.

(3) " Exxon Nuclear Critical Power Methodology for Boiling Water Reactors",XN-NF-524(P), Exxon "uclear Company Inc., Richland, Washington99352, November 1979.

(4) "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix KHeatup Option", XN-CC-33(A), Revision 1, Exxon Nuclear CompanyInc., Richland, Washington 99352, November 1975.

(5) " Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies",XN-NF-79-59(P), Exxon Nuclear Company Inc., Richland, Washington99352, 1979.

(6) "Dresden-3 Cycle 8 Plant Transient Analysis Report", XN-NF-81-78,Revision 1, Exxon Nuclear Company Inc., Richland, Washington 99352,December 1981.

I

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

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A-1 XN-NF-82-84 (NP) |Revision 1

APPENDIX A

DRESDEN UNIT 2, CYCLE 9

SAFETY LIMIT CALCULATION PARAMETERS,

INPUT VALUES, AND UNCERTAINTIES

!

A.1 REACTOR SYSTEM UNCERTAINTIES

The reactor system uncertainties used in the Dresden Unit 2 Cycle 9

safety limit calculation are the generic vaines listed in Table 5.1 of XN- |

NF-524(P)(3).

A.2 FUEL RELATED UNCERTAINTIES

Ful related uncertainties used in the Dresden Unit 2 Cycle 9 safety

limit calculation are listed in Table A-1. The values listed in Table A-1

are for Dresden Unit 2 Cycle 9 with the exception of the XN-3 correlation

uncertainty, which is generic.

A.3 NOMINAL INPUT PARAMETER VALUES

Nominal values of input parameters used in the Dresden Unit 2 Cycle 9

safety limit calculation are listed in Table A-2.

A.3.1 RADIAL POWER HISTOGRAM

The radial power histogram used in the Dresden Unit 2 Cycle 9

safety limit calculation is given in Figure A-1. The radial power

histogram was chosen from a representative group of histograms. The

histogram was then biased in a manner which would produce a worse (larger)

value of the predicted safety limit. The peak value for the histogram was

chosen such that the limiting bundle MCPR would conservatively remain

greater than the expected MCPR operating limit under steady-state, full-

power, full-flow conditions.

._ _ _ _ -

A-2 XN-NF-82-84(NP)Revision 1

A.3.2 LOCAL PEAKING DISTRIBUTION

The local peaking distribution used in the Dresden Unit 2

Cycle 9 safety limit calculation is shown in Figure A-2. The local peaking

distribution was chosen from the predicted distributions covering the

range of Dresden Unit 2 Cycle 9 exposures. The chosen distribution was

used because it was found to produce the worst (largest) value of the

safety limit of the group of distributions.

A.3.3 AXIAL POWER DISTRIBUTION

The axial power distribution used in the Dresden Unit 2 Cycle,

| 9 safety limit calculation was:l

FA (X/L) = 0.30 + 1.10 sin (gX/L)

where X/L = relative axial position. This axial power distribution was

chosen because it is conservative with respect *.o the predicted axial power

distributions of MCPR limiting bundles.

A.4 SAFETY LIMIT RESULTS~

The final Dresden Unit 2 Cycle 9 safety limit calcu1ation used 500

Monte Carlo trials. The MCPR of the safety limit calculation using the

nominal input parameters was 1.05 or less for all fuel types. With those

conditions, the number of rods in the core which are expected to avoid

boiling transition is greater than 99.9%. Thus, a safety limit of 1.05 for!

| Dresden Unit 2 Cycle 9 satisfies the requirement that at least 99.9% of the

rods in the core must be expected to avoid boiling transition when the|

|reactor is at the safety limit.

!

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- _ _ _ _ , _ . . _ , _ . _ . - . . _ __. _ _ . _ . _ . .

..


Table A-1 Dresden Unit 2 Cycle 9 Safety Limit'

Fuel Related Uncertainties

Parameter Standard Deviation Assumed Probability(% of Nominal) Distribution Type

XN-3 correlation 4.1 Normal

Assembly Radial Peaking 5.18 NormalFactor

Rod Local Peaking Factor 2.46 Normal

Assembly Flow Rate 2.8 Normal

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4

Table A-2 Dresden Unit 2 Cycle 9 Safety Limit Nominal Input Parameters,

,

Parameter Value

i

Core Pressure 1035 psia

Core Power 3277 MW

Core Inlet Enthalpy 521.8 BTU /lbm

Total Core Flow 98.0 Mlbm/hr

Feedwater Temperature 3200F

Feedwater Flow Rate 12.4 Mlbm/hr

Hydraulic Demand Curve *

G = 1.540 + (-8.851 x 10-2) x LHGR

+ (1.908 x 10-3) x LHGR2 (8x8 fuel)

'2where G = Assembly Mass Flux [Mlbm/ft -hr]

LNGR = Assemblypower[kw/ft] -

___________________________

Reference 2, Section 3.3*

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1A-6 XN-NF-82-84 (NP).

Revision 1.

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...._____.....____..........._....__..___....___.........

. . . . . . .. . . . . . .

: L : ML : ML : M : M : " : ML : ML :: 1.06 : 1.02 : 3.92 1.N : 1."6 : 1.08 : 0.92 : 1 01 :

*

. . . . . . . ... . . . . . . .

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: ML : ML* : M : H : H : ML* : M : PL :| 1.38 : 0.90 : 1 03 : 1.00 : 0.98 : 0.73 : 1. 3 : 3.92 :

-

. . . . . . .

. . . . . . .

_____._____........__ ..___......____....._______......... . . . .. . . . .

: ML : # : H : H H : H : ML* : M :*

: 1.03 : 1.09 : 1.00 : 0.94 : 0 95 : 0.96 : 0.73 : 1.08 :. . . . . . . . .. . . . . . . . .

..___________........... _____...___..__......_..___ ..... . . .. . . .

: ML : M : H : H : W : H : H : M :: 1 01 : 1.06 : 0.97 : 0.94 : 0.93 : 0.95 : 0.9S : 1.06 :. . . . . . .. . . . . . .

_.........______...___............._______..... _.......

. . . . .. . . . .

: ML : M : H : H : H : H H M :-* *

: 1.03 : 1.06 0.99 : 3.94 : 0 94 : 0.94 : 1.00 : 1.08 :*

. . . . . . . . .. . . . . . . . .

_____. ....___......._____..__ ...._______......_____..... . . . . . .. . . . . . .

: ML : ML : M : H : H : H : M : ML :: 1 08 : 0.43 : 1.33 : 0.99 : 3.97 : 1.00 : 1.03 : 0.92 :. . . . . . . .. . . . . . . .

..____........_______........__...__________. __________

. . . . . . .. . . . . . .

: L : "La : ML : M : M : M : ML* : ML :: 1 07 : 1 01 : 0.93 : 1.~8 : 1.06 : 1.09 : 0.93 : 1.02 :. . . . . . .. . . . . . .

y ..__...........___.....__... ........ ....... ........___I ..

. . . . .. . . . . .

D : LL : L : ML : ML : ML : ML : ML : L :C : 1 01 : 1 27 : 1.98 : 1.03 : 1.01 : 1.03 : 1 08 : 1.06 :

. . . . . . . .. . . . . . .

..............._.. _______..__...__........... ____.._ ..

I WIDE

Figure A-2 Dresden-2 Cycle 9 Safety Limit Local Peaking

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B-1 XN-NF-82-84(NP)Revision 1

APPENDIX B

DRESDEN UNIT 2, CYCLE 9 EXPERIMENTAL DESIGN

The experimental design used in the construction of the response surface

is provided in the attached Table B-1. This design is a Box-Behnken type for

N=4 as described in XN-NF-81-22(P). The variables are defined as follows:- -

XI -= -.

X: -

= :

X3 -= -

X4 --

The coded values (+2, 0, -2) for the variables are as described in XN-NF-81-

22(P), except that the standard deviations used are specific to Dresden Unit

2 hee Section 3.2.1 of main text). The sign of the coded variable values for

the first three variables was chosen such that a positive value would increase

the observed CPR. The opposite applies to the fourth (X4) variable.

.

_- - - _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ . . _

_...

- _ - . - _____

B-2Revision 1 (NP)XN-NF-82-84

Table B-1 ACPR Experimental Design

Coded Values of Predictor Variables )-

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XN-NF-82-84(NP)Revision 1

Issue Date: 12/07/82

PLANT TRANSIENT ANALYSIS FOR ,

DRESDEN UNIT 2 CYCLE 9

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Distribution

J. C. Chandler

G. C. Cooke

N. F. Fausz

R. H. Kelley

L. C. O'Malley/ Edison (60)

Document Control (5)

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nonproprietary version of revision 1 to 'plant transient

Documents