ML20204F631

Errata to Rev 1 to Reload Core Analysis Methodology Overview

ML20204F631
Person / Time
Site:	Fort Calhoun
Issue date:	03/18/1987
From:	OMAHA PUBLIC POWER DISTRICT
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ML19292G995	List: ML20204F393 ML20204F478 ML20204F585 ML20204F631
References
OPPD-NA-8303-NP-R01E, OPPD-NA-8303-NP-R1E, TAC-63967, NUDOCS 8703260232
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4.0 TRANSIENT AND ACCIDENT ANALYSIS MODELS (Continued) 4.1 Pl. int Simulation Model The District utilizes the CESEC digital computer code, References 4-1 through 4-10, to provide the simulation of the Fort Calhoun Station nuclear steam supply system. The program calculates the plant response to non-LOCA initiating events for a wide range of operating .

conditions. The information presented in Reference 4-9 supercedes information provided in References 4-1 through 4-8. Additional information on the model is provided in Reference 4-10. The CESEC program, which numerically integrates one dimensional mass and energy conservation equations, assumes a node / flow-path network to model the NSSS. The primary system components considered in the code include the reactor vessel, the reactor core, the primary coolant loops, the pressurizer, the steam generators and the reactor coolant pumps. The secondary system components include the secondary side of the steam generators, the main steam system, the feedwater system and the various steam control valves. In addition, the program models some of the control and plant protection systems.

The code self initializes for any given, but constant, set of reactor power level, reactor coolant flow rate and steam generator power sharing. During the transient calculations, the time rate of change in the system pressure and enthalpy are obtained from solution of the conservation equations. These derivatives are then numerically inte-grated in time under the assumption of thermal equilibrium to give the system pressure and nodal enthalpies. The fluid states recognized by the code are subcooled and saturated; superheating is allowed in the pressurizer. Fluid in the reactor coolant system is assumed to be 8703260232 870318 PDR ADOCK 05000285 p PDR O

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5.0 TRANSIENTANDALCfDENTANALYSISMETHODS(Continued) 5.1 CEA Withdrawal (Continued) 5.1.1 Definition of the Event (Continued)

Any controlled or unplanned withdrawals of the CEA's results in a positive reactivity addition which causes the core .

power, core average heat flux and reactor coolant system temperature and pressure to rise and in turn decrease the DNB and Linear Heat Rate (LHR) margins. The pressure increase, if large enough, activates the pressurizer sprays which mitigate the pressure rise. In the presence of a positive Moderator Temperature Coefficient (MTC) of reac-tivity, the temperature increase results in an additional positive reactivity addition further decreasing the margin to the DNB and LHR limits.

O Withdrawal of the CEA's causes the axial power distribution to shift to the top of the care. The associated increase in the axial peak is partially compensated by the corresponding decrease in the integrated radial peaking factor. The magnitude of the 3-D peak change depends primarily on the initial CEA configuration and axial power distribution.

The withdrawal of the CEA's causes the neutron flux as measured by the excore detectors to be decalibrated due to CEA motion, i.e., rod shadowing effects. This decalibration of excore detectors, however, is partially compensated by neutron attenuation rising from moderator density changes (i.e. , temperature shadowing effects).

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Table 5.5.4-1 KEY PARAMETERS ASSUMED IN THE LL/ISG EVENT Parameter Units Value ,

l Initial Core Power MW 700 to 1500t t

Initial Core Inlet *F Maximum allowedt '

Temperature by Tech. Specs.

Initial Reactor Coolant psia Minimum allowedt .

System Pressure by Tech. Specs.

Moderator Temperature *10~%o/ F Coefficient i Fuel Temperature *10~ %o/"F l Coefficient Core Average Hgap BTU /hr-ftz- F Maximum value predict-  :

ed during core life.  !

Initial Core Mass *1061bm/hr. Best estimate flowt Flow Rate Scram Reactivity Worth %Ap Minimum predicted during core life.

tFor DNBR calculations, effects of uncertainties on these parameters were combined statistically.

i O

OPPD-NA-8303-NP Rev. 01 Page 42 of 136

, 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued)

(G 5.5 Asymmetric Steam Generator Event (Continued) 5.5.6 Analysis Results and 10 CFR 50.59 Criteria The results of the analysis for the LL/1SG event are dis-cussed in Section 7 of Reference 5-2.

The results for Fort Calhoun Station are expected to be similar. The criteria of 10 CFR 50.59 are satisfied if the required overpower margin calculated for the LL/1SG event is less than the overpower margin being maintained by the current Technical Specifications.

5.6 Excess Load Incident O

G 5.6.1 Definition of Event An excess load transient is defined as any rapid increase in the steam generator steam flow other than a steam line break. Such a rapid increase in steam flow results in a power mismatch between the reactor core and the steam generator load demand. In addition, there is a decrease in the reactor coolant temperature and pressure. Under these conditions the negative moderator temperature coefficient reactivity causes an increase in core power.

The rapid opening of the turbine admission valves or the steam dump bypass to the condenser causes an excess load event. Turbine valves are not sized to accommodate steam flow for powers much in excess of 1500 MWt. The steam dump valves and steam bypass valves to the condenser are sized to accommodate 33% and 5%, respectively, of the steam flow at O 1500 MW. Therefore, the following load increase incidents V are examined:

OPPD-NA-8303-NP Rev. 01 Page 43 of 136

5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.6 ExcessLoadIncident(Continued) 5.6.1 Definition of Event (Continued)

C. (Continued)

(540*F). The maximum error that can be introduced in the referenced temperature setting is limited to 17'F since a narrow range instrument is used for this purpose. Reducing the dump valve controller reference setting from 532* to 515 would result in a partial opening of the valves but as soon as the reactor coolant temperature dropped to 518'F the valves would again be completely closed.

D. Opening the dump and bypass valves at hot standby due to steam dump controller malfunction: The most severe incident at hot standby would occur in the event the steam dump valve controller yields an incorrect signal and causes the steam dump and bypass valves to open completely. This case is considered to be much less probable than case C above but represents the most limiting event under hot standby conditions.

The possible RPS trips that might be encountered during this event are:

1. Variable high power trip (VHPT).

2. TM/LP trip.

3. Low steam generator water level trip.

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C 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.6 Excess Load Incident (Continued) 5.6.4 Key Parameters and Analysis Assumptions As discussed in Section 5 of Reference 5-2, sensitivity studies performed by CE have demonstrated that the maximum calculated [ ] for the excess load event occurs for the [

]athotfullpowerconditions. District sensitivity studies show similar results. Therefore, only the hot full power case is analyzed. The key parameters used in the analysis of the excess load event are given in Table 5.6.4-1. The remaining assumptions are the same as those discussed in Reference 5-2.

5.6.5 Analysis Method The steps used for determining the [ ]valueand calculating the largest [ ] for all excess load events which rely on the TM/LP trip for DNBR protection are given in Section 5 of Reference 5-2. The minimum transient DNBR value for excess load events protected by the Variable High Power Trip is calculated using the procedure discussed in the same Section.

The PLHR is calculated by obtaining the core average linear heat rate at time of peak core power and multiplying it by the appropriate peaking factors and associated uncertain-ties.

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Table 5.6.4-1 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD EVENT ANALYSIS Parameter Units Value Initial Core Power MW t

1500t Initial Core In- Maximum allowedt let Temperature 'F At Power by Tech. Specs.

Initial Reactor Coolant psia Minimum allowedt .

System Pressure by Tech. Specs.

Initial Core Mass *1061bm/hr. Minimum allowedt Flow Rate by Tech. Specs.

CEA Drop Time sec. Maximum allowed by Tech. Specs.

Scram Reactivity %Ao Minimum predicted Worth during core life.

Moderator Temperature *10 4ap/*F Negative values up to Coefficient the most negative value allowed by Tech. Specs, O

t For DNBR calculations, effects of uncertainties on these parameters were combined statistically.

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V 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.6 Excess Load Incident (Continued) 5.6.6 Analysis Results and 10 CFR 50.59 Criteria The results of the excess load analysis are similar to those presented in Section 5 of Reference 5-2. The criteria of 10 l CFR 50.59 are met if the [ ]islessthanor equal to the value used in the current TM/LP trip equation.

5.6.7 Conservatism of F.esults The following points demonstrate the conservatism of the overall results for the excess load event:

1. Field measurements demonstrate that the CEA magnetic O clutch decay time is less than that assumed in the i

b analysis.

l 2. The actual scram worths are higher than those in the analysis.

3. Where the most negative MTC is used, the value is more negative than that measured during plant operation.

4. The actual Doppler reactivity is more negative than assumed in the analysis.

5.

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Q kJ 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHODS (Continued) 5.6 Excess Load Incident (Continued) 5.6.7 Conservatism of Results (Continued)

6. Field data demonstrates that the actual CEA drop time is less than that assumed in the analysis.

7. The conservatism of the [ ] is dis-cussed in Section 5 of Reference 5-2.

5.7 RCS Depressurization 5.7.1 Definition of Event I

The RCS depressurization event is characterized by a rapid decrease in the primary system pressure caused by either the inadvertent opening of both power operated relief valves (PORVs) or the inadvertent opening of a single primary safety valve operating at rated thermal power. Following the initiation of the event, steam is discharged from the pressurizer steam space to the quench tank where it is condensed and stored. To compensate for the decreasing pressure the water in the pressurizer flashes to steam and the proportional heaters increase the heat added to the water in the pressurizer in an attempt to maintain pressure. .

During this time the pressurizer level also begins to decrease causing the letdown control valves to close and additional charging pumps to start so as to maintain level.

As pressure continues to drop, the backup heaters energize to further assist in maintaining primary pressure. A reactor trip is initiated by the TM/LP trip to prevent exceeding the DNDR SAFDL.

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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued)

L 5.8 MainSteamLineBreakAccident(Continued) I a 5.8.5 Analysis Method The analysis of the main steam line break accident is  :

performed using CESEC which models neutron kinetics with fuel and moderator temperature feedback, the reactor pro-tective system, the reactor coolant system, the steam l generators and the main steam and feedwater systems.

5.8.6 Analysis Results and 10_CFR 50.59 Criteria f 1

i i The results of the analysis for the Fort Calhoun steam line i

! break event are discussed in Section 14.12 of the 1983 l update of the Fort Calhoun Station l' nit No.1 USAR. The j criteria of 10 CFR 50.59 are met if the calculated return-to-power is less than the return-to-power reported for the  !

j Cycle 1 analysis, using the current Technical Specification  !

I j limit on shutdown margin and moderator temperature coef- l

! ficient.

! i i

5.8.7 Conservatism of Results i

Conservatism is added to the analysis by inclusion of uncertainties in moderator and fuel temperature coefficients of reactivity, by taking no credit for void reactivity feedback, by taking credit for only 1 IIPSI pump and by

]

taking no credit for the stuck CEA worth, i

i I

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OPPD-NA-8303-NP i Rev. 01 l l

Page 62 of 136 j t  ;

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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued) 5.10 CEA Ejection Accident 5.10.1 Definition of Event A CEA ejection accident is defined as a mechanical failure of a control rod mechanical pressure housing such that the coolant system pressure would eject the CEA and the drive shaft to a fully withdr wn position. The consequences of this mechanical failure 1. a rapid reactivity insertion which when combined with an adverse core power distribution potentially leads to localized fuel damage. The CEA ejection accident is the most rapid reactivity insertion that can be reasonably postulated. The resultant core and thermal power excursion is limited primarily by the Doppler reactivity effect of the increased fuel temperatures and is terminated by reactor trip of the rernaining CEA's activated by the high power trip or variable high power trip.

5.10.2. Analysis Criteria The CEA ejection event is classified as a postulated accident. The design and limiting criteria are:

1. Fuel cladding and enthalpy thresholds (Reference 5-5) are:

Clad Damage Threshold Total Average Enthalpy = 200 cal / gram Centerline Melting Threshold Total Centerline Enthalpy = 250 cal / gram Fully Molten Centerline Threshold Total Centerline Enthalpy = 310 cal / gram OPPD-NA-8303-NP Rev. 01 Page 68 of 136

i 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued)

C 5.10 CEA Ejection Accident (Continued) 5.10.5 Analysis Results and 10 CFR 50.59 Criteria The results of the CEA Ejection Analysis are reported in Section 14.13 of the Fort Calhoun Station Unit No.1 USAR.

Criteria of 10 CFR 50.59 ere satisfied if fuel failures are less than those assumed for input to the Radiological Consequences portion of the analysis. Cycle 10, as an example, utilized 1% fuel failure.

5.10.6 Conservatism of Results The major area of conservatism is the calculation method used to obtain the ejected CEA worth and the ejected radial peak. The ejected worth and the ejected radial peak are calculated without any credit for Doppler or Xenon feedback, in addition, the hot full power ejected worth and ejected peak are calculated assuming the no-load temperature of 532 F. The lower temperature is more adverse since this causes a power role to the core periphery which also happens to be the location of the ejected CEA. Also, the ejected worth is calculated assuming the CEA'S are fully inserted for hot full power case regardless of P0ll. Thus, the ejected worth is conservative.

5.11 Loss of Coolant Accident The District does not perform the Loss of Coolant Accident Analysis.

The large and small break loss of coolant analyses were performed by Combustion Engineering (CE). The large break topical is mentioned in Reference 5-8. The small break analysis shnws the closest approach to

( ) the 10CFR50.46 criteria for ECCS analysis. The actual difference is OPPD-NA-8303-NP Rev. 01 Page 70 of 136 A

(a 5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued) 5.12 Loss of Load to Both Steam Generators Event (Continued) 5.12.2 Analysis Criteria (Continued) a) The peak RCS pressure does not exceed 2750 psia (110%

of design pressure).

l .

b) The transient minimum DNBR is greater than the 95/95 confidence interval limit for the CE-1 correction limit.

a. ) The Peak Linear Heat Generation Rate (PLHGR) does not j exceed 22 kw/ft.

Criteria b. and c. are not of major concern since DNPR increases during the event and the PLHGR margin required is much less limiting than other A00's. Therefore, criterion

a. is the main concern in analyzing this event. The loss of load to both steam generators event is the limiting A00 event with respect to peak RCS pressure.

5.12.3 Objectives of the Analysis The objective of the analysis is to demonstrate, for modifications to the plant which potentially degrade RCS heat removal capability (including steam generator plugging) that the peak RCS pressure stays within 110% of the design pressure in accordance with Section III of the ASME Pressure Vessel Code. This objective is achieved if the peak RCS pressure does not exceed 2750 psia, 4

e OPPD-NA-8303-NP Rev 01 Page 72 of 136

Table 5.12.4-1 KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD TO BOTH STEAM GENERATORS ANALYSIS Parameter Units Value Initial Core Power Level MW t

1530(102%)

Initial Core Inlet Maximum allowed Temperature *F by Tech. Specs.

Initial RCS Pressure psia Minimum allowed by Tech. Specs Initial Steam Generator psia Minimum value cor-Pressure responding to core inlet temperature operating range.

Initial Core Mass *106 lbm/hr. Minimum allowed Flow Rate by Tech. Specs.

Moderator Temperature

• 10~%o/'F Most positive Coefficient allowed by Tech, Specs.

f)

G' Fuel Temperature *10~%p/'F Least negative Coefficient predicted during core life.

Fuel Temperature Coefficient Multiplier 0.85 CEA Drop Time sec. Maximum allowed by Tech. Specs.

Scram Reactivity Worth Minimum predicted during core lifetime Scram Reactivity Consistent with most Curve ositive axial shape p(bottompeaked) allowed by Tech. Specs.

Core Average H BTU /hr-Fte "F Haximum predicted UdP during core lifetime.

Kinetics Parameters EOC parameters (minimumabsolute 8).

O. RPS Rtspense Time sec. 1.4 OPPD-NA-8303-NP Rev. 01  :

Page 74 of 136

5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued) 5.12 Loss of Load to Both Steam Generators Event (Continued) 5.12.7 Conservatism of Results (Continued)

6. The maximum pressurizer safety valve capacities are assumed to be 90% of the ASME rated values.

7. A one percent pressure uncertainty is applied to the primary and secondary safety valve setpoints, i.e., a 1.01 multiplier.

5.13 Loss of Feedwater Flow Event 5.13.1 Definition of Event A total loss of main feedwater flow event is defined as a loss of feedwater flow when operating at power without a corresponding reduction in steam flow from the steam generators. The most likely causes for this event are the loss of all feedwater or condensate pumps or the inadvertent closure of either the main feedwater regulating valves or the feedwater isolation valves due to a feedwater controller malfunction or manual positioning by the operator. The result of this mismatch in which turbine demand remains at 100%, is a reduction of the steam generator liquid inventories and a degrading RCS heat removal capability.

As the heat removal capability is lost, through decreasing steam generatorinventories(i.e., levels)theRCStemperaturesand pressure increase. Normally the event would be terminated by a reactor trip on low steam generator level. Since no credit is taken in the analysis for the steam generator low level trip, a high pressurizer trip eventually results.

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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued) 5.13 Loss of Feedwater Flow Event (Continued) 5.13.1 Definition of Event (Continued)

Automatic actuation of the auxiliary feedwater (AFW) system will also eventually occur (after reactor trip) if either main feedwater is not restored or manual actuation of the AFW system is not performed by the operator. The AFW system actuation ensures the maintenance of a secondary heat sink.

5.13.2 Analysis Criteria The loss of feedwater flow event is classified as an Anticipated Operational Occurrence (A00) for which the following criteria must be met:

O a. The peak RCS pressure does not exceed 2750 psia (110%

ofdesignpressure).

b. The transient minimum DNBR is greater than the 95/95 confidence interval limit for the CE-1 correlation limit.

c. ThePeakLinearHeatGenerationRate(PLHGR)doesnot exceed 22 kw/ft.

Criteria b. and c. are not of major concern because DNBR does not decrease below the initial steady state value and the PLHGR margin required is much less limiting than other A00's. Therefore, only criterion a requires reevaluation should plant modifications (such as steam generator tube O

OPPD-NA 8303-NP Rev. 01 Page 77 of 136

5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued) 5.13 Loss of Feedwater Flow Event (Continued) 5.13.2 Analysis Criteria (Continued) plugging) be made which result in degraded secondary heat transfer capability beyond that of this event. For Fort Calhoun Station, this event is bounded by the loss of load incident.

5.13.3 Objectives of the Analysis The objective of this analysis is to demonstrate, for plant modifications which potentially degrade RCS heat removal capability (includingsetamgeneratortubeplugging),that the peak RCS pressure stays within 110% of the design pressure in accordance with Section 111 of the ASME Pressure Vessel Code. This objective is achieved if the peak RCS pressure does not exceed 2750 psia.

5.13.4 Key Parameters and Analysis Assumptions The key parameters used in the loss of feedwater flow event are given in Table 5.13.4 1. Assumptions in the analysis to maximize heat up of the RCS and consequently the peak RCS pressure include:

1. The event is initiated by an instantaneous loss of main feedwater. No credit is taken for the low steam generator level trip.

2. The steam dump and bypass system is assurred to be in MANUAL (i.e., inoperative).

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5.0 TRANSIENT AND ACCIDENT ANALYSIS METHOD (Continued) 5.13 Loss of Feedwater Flow Event (Continued) 5.13.4 Key Parameters and Analysis Assumptions (Continued)

3. The pressurizer pressure control system is in MANUAL (i.e.,PORV'sandspraysareinoperable).

4. The pressurizer level control system is in MANUAL with maximum charging and zero letdown flows.  ;

5. The rod block system is assumed to prevent rod motion  ;

(other than scram) during the transient.

5.13.5 Analysis Method The analysis methods used by the District to analyze a loss of main feedwater flow event consists of using the CESEC computer code to simulate the event, utilizing the analysis assumptionlistedinSection5.13.4(above)asinput,and extracting the peak RCS pressure for comparison with the 2750 psia upset limit.

5.13.6 Analysis Results and 10 CFR 50.59 Criteria The results of the loss of feedwater flow are contained in the Fort Calhoun Station Unit No. 1 USAR. The criteria of 10 CFR 50.59 are met, if the peak RCS pressure is less than the value reported in Section 14.10.1 of the USAR.

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Page 79 of 136

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( Table 5.13.4-1 KEY PARAMETERS ASSUMED IN THE LOSS OF FEEDWATER FLOW ANALYSIS Parameter Units Value Initial Core Power Level MW t

1530 (102%)

Initial Core Inlet Maximum allowed Temperature *F by Tech. Specs.

Initial RCS Pressure psia Minimum allowed by Tech. Specs Initial Steam Generator psia Minimum value cor-Pressure ' responding to core inlet temperature operating range.

Initial Core Mass *106 lbm/hr. Minimum allowed Flow Rate by Tech. Specs.

Moderator Temperature *10~%o /* F Most positive Coefficient allowed by Tech.

Specs.

Fuel Temperature *10~%o/*F Least negative Coefficient predicted during core life.

Fuel Temperature Coefficient Multiplier 0.85 CEA Drop Time sec. Maximum allowed by Tech. Specs.

Scram Reactivity Worth Minimum predicted during core lifetime Scram Reactivity Consistent with most Curve ositive axial shape p(bottompeaked) allowed by Tech. Specs.

Core Average 11 BTU /hr-Ftr.oF Haximum predicted 98P during core lifetime.

Kinetics Parameters E0C parameters (minimum absolute 8).

RPS Response Time sec. 1.4 OPPD-NA-0303-NP Rev. 01 Page 80 of 136

[v ') 5.0 TRANSIENTANDACCIDENTANALYSISMETHOD(Continued) 5.13 Loss of Feedwater Flow Event (Continued) 5.13.6 Analysis Results and 10 CFR 50.59 Criteria The results of the loss of feedwater flow are contained in the Fort Calhoun Station Unit No. 1 USAR. The criteria of 10 CFR 50.59 are met, if the peak RCS pressure is less than the value reported in Section 14.10.1 of the USAR.

5.13.7 Conservatisms of Results

1. Field measurements demonstrate that the CEA magnetic clutch decay time is less than that assumed in the analysis.

O 2. The actual scram worths are greater than those assumed in the analysis.

3. The actual MTC is more negative during power operation than assumed in the analysis.

4. The steam dump and bypass system and the pressurizer pressure control system (PORV's and sprays) are operated in the AUTO mode rather than the MANUAL mode as assumed in the analysis.

5. Actual secondary pressure is higher which results ia earlier secondary safety valve opening and earlier alleviation of the primary system temperature and pressure rises.

6. No credit is taken for a steam generator low level trip.

v OPPD-NA-8303-NP Rev. 01 Page 81 of 136