ML20002A277

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Suppl 1 to Application for Amend of License DPR-53 to Allow Fifth Cycle Operation
ML20002A277
Person / Time
Site: Calvert Cliffs Constellation icon.png
Issue date: 11/04/1980
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20002A276 List:
References
NUDOCS 8011050510
Download: ML20002A277 (82)
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Text

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EllCLOSURE 1 SUPPLE' . ~4f 1 TO CALVERT CLIFFS Uf1IT I CYCLE 5 REFUELIl1G LICEISE Af1EliD'tEt1T 80110505/O

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l SUPPLEMEilT 1 TO CALVERT CLIFFS UilIT I l CYCLE 5 REFUELIrlG LICENSE AMEflDMENT Table of Contents Section

1. Introduction and Summary
2. Operating History of the Reference Cycle
3. General Description
4. Fuel System Design
5. fluclear Design
6. Thermal-Hydraulic Design
7. Transient Analysis
8. ECCS Analysis
9. Technical Specifications
10. Startup Tes ting
11. References Appendices A. Asymmetric Steam Generator Transient Pro-tection Trip Function B. Method for Calculating Space-Time Scram Reactivities C. Description of Model Used to Simulate NSSS Behavior During Steam Line Rupture Event D. Prototype Assembly Description E. Description of Modified Assembli .s

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1. INTRODUCTION AND

SUMMARY

No change from original request except that the maximum E0C4 burnup l window has been increased from 11,600 MWD /T to 11,800 MWD /T.

i l 2. OPERATING HISTORY OF THE REFERENCE CYCLE

-ilo change from original request except 'that the maximum E0C4 burnup window has been increased- from 11,600 MWD /T to 11,800 MWD /T.

3. GENERAL DESCRIPTION 1 No change from original request except that the maximum E0C4 burnup 1 window has been increased from 11,600 MWD /T to 11,800 MWD /T.
4. FUEL SYSTEM DESIGN No change from original request.

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5. NUCLEAR DESIGN No change from original request except that the maximum E004 burnuo window has been increased from 11,600 MWD /T.to 11,800 MWD /T, and

] Section 5.3 is revised to eliminate reference to the' FIESTA code.

5.3 _ NUCLEAR DESIGN METHODOLOGY (Below information supplements or modifies original request).

The analyses have been performed with the same methodologies used for the reference analyses.

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l 6.0 THERMAL HYDRAULIC DESIGN (Belcw Information Modifies Original Request).

6.1 DNBR Analysis Steady state DNBR analyses of Cycle 5 at the rated power level of

?700 MWT have been performed using the TORC computer code described n Reference 1, and the CE-1 critical heat flux correlation described in Reference 2.

As was done for Cycle 4, TORC /CE-1 was used in this cycle to develop the

" design thermal margin model", described generically in Reference 3. TORC was used for all A00s and postulated accidents reanalyzed for Cycle 5.

Table 6-1 contains a list of pertinent thermal-hydraulic design parameters used for both safety analyses and for generating reactor protective system setpoint information. Also, the calculational factors (engineering) heat flux factor, engineering factor on hot channel heat input and rod pitch, bowing and clad diameter factor) are listed in Table 6-1.

Investigations have been made to ascertain the effect of the CEA guide tube wear problem and the sleeving repair on DNBR margins as established by this type of analysis. The findings were reported to the NRC in Reference 4 which concluded that the wear problem and the sleeving repair do not adversely affect DNBR margin.

6.2 Effects of Fuel Rod Bowing on DNBR Margin Effects of fuel rod bowing on DNBR margin have been incorporated in the safety and setpoint analyses in the same manner as discussed in Reference S. This reference contains penalties on minimum DNBR due to fuel rod bowing as a function of burnup using NRC guidelines contained in Reference 6.

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TABLE 6-1

-Calvert Cliffs Unit 1 Thermal-Hydraulic Parameters at Fu;i Power Reference General Characteristics Unit Cycle 4 Cycle 5 Total Heat Output (core only) MWT 2700 '2700 106 Btu /hr 9215 9215' Fraction of. Heat Generated in .975 .975 Fuel Rod Primary System Pressure Nominal. psia- 2250 2250 Minimum in steady state psia 2200 2200 Maximum in steady state psia 2300 2300 Inlet Temperature F 550 550 Total Reactor Coolant Flow gpm 370,000 .370,000 (steady state) 106 lb/hr 139.0 139.0

. Coolant Flow Through Core 106 lb/hr 135.3 133.9*

Hydraulic Diameter ft 0.044 0.044

-(nominal channel)

Average Mass Velocity 106 lb/hr-ft 2.53 2.51 Pressure Drop Across Core psi 10.6 10.4 (minimum steady state flow irreversible op over entire fuel assembly)

Total Pressure Drop Across Vessel psi 32.6 32.4 (based on nominal dimensions and minimum steady state flow)

Core Average Heat Flux (accounts Btu /hr-ft 181,200 186,435**

for above fraction of heat generated in fuel rod and axial densification factor) 2 Total Heat Transfer Area (Accounts ft 49,600 48,192**

for axial densification factor)

Film Coefficient' at Average Conditions Btu /hr-ft2 - F 5815 5765 Average Film Temperature Difference F 31 32 Average Linear l Heat Rate of kw/ft 6.05 6.45**' l Undensifled Fuel Rod (accounts  !

for above fraction of. heat l generated in fuel rod)-

Average Core Enthalpy-Rise Btu /lb 68 68.8 -

. Maximum Clad -Surface Temperature - F 657 657 I J

Table 6-1 (cont.)

Calculational Factors Reference Cycle 4 Cycle 5 Engineering Heat Flux Factor 1.03 1.03 Engineering Factor on Hot Channel Heat 1.03 1.02 (as-built)

Input Rod Pitch, Bowing and Clad Diameter 1.065 1.065 Fuel Densification Factor (axial) 1.01 1.01 NOTES All flow-related parameters are calculated at design inlet temperature, nominal primary system pressure.

  • Based on a 3.7% generic core bypass flow. Actual Cycle 5 coolant flow through the core is 135.3x100 lb/hr.
    • Based on generic value of 1100 shims. Actual Cycle 5 number of shims is 416.

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7.0 TRANSIENT ANALYSIS The purpose of ~this section is to present the supplemental results of the 1

Baltimore Gas & Electric Calvert Cliffs Unit 1, Cycle 5 Non-LOCA safety

- analysis' at 2700 MWt without the use of the SCO, FIESTA and CETOP programs.

The information presented in this section supplements or modifies the original request. The Design Bases Events (DBEs) considered in the safety analyses are listed in Table 7-1, Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to

' melt fuel design ilmits are presnted in Table 7-2.

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l TABLE 7-1 CALVERT CLIFFS UNIT 1 CYCLE 5 DESIGN BASIS EVENTS CONSIDERED IN NON-LOCA SAFETY ANALYSIS Analysis Status 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits:

7.1.1 Sequential CEA Group Withdrawal Reanalyzed 7.1.2 Boron Dilution No change from original request except for refuel-ing shutdown margin 7.1.3 Startup of an Inactive Reactor Coolant Pump No change from original

request I

7.1.4 Excess Load Reanalyzed 7.1.5 Loss of Load Reanalyzed 7.1.6 Loss of Feedwater Flow No change from original request 7.1. 7. Excess heat Removal due to Feedwater Malfunction No change from original request 7.1.8 Reactor Coolant System Depressurization No change from original request 7.2 Anticipated Operational Occurrerces for which RPS trips and/or sufficient initial steady state thermal margin, -

maintained by the LCOs, are necessary to prevent exceeding the acceptable limits

1 i Reanalyzed 7.2.1 Loss of Coolant Flow l 7.2.2 Loss of AC Power No change from original

request 7.2.3 Full Length CEA Drop Reanalyzed 7.2.4 Transients Resultino From the Malfunction of Reanalyzed One Steam Generator 2

TABLE 7-1 (continued) 7.3 Postulated Accidents:

, 7.3.1 CEA Ejection No cnange from original request except for results ,

of Table 7.3.1-2 7.3.2 Steam Line Rupture No change from original request 1

7.3.3 Steam Generator Tube Rupture No change fro 1 original j request l 7.3.4 Seized Rotor Reanalyzed 1

Requires Low Flow trip.

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, Requires trip on high differential steam generator pressure; this trip function .

is described in Appendix A of tne original request.

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E TABLE 73 CALVERT CLIFFS UNIT 1 CYCLE 5 CORE PARAt1ETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle Values Cycle 5 Physics Parameters Units (Cycle 4) Values Radial Peaking Factors For DNB Margin Analyses (Fr)

Unrodded Region 1.58 1.62

! Bank 5 Inserted 1.71 1.78 For Planar Radial Component (F y) of 3-D Peak (CTM Limit Analyses')

Unrodded Region 1.66 1.62 1,79 1.78 Bank 5 Inserted Maximum Augmentation Factor 1.069 1.055 Moderator Temperature Coefficient 10~4ao/ F -2.5 + +.5 -2.5*+ +.5 Shutdown flargin (Value assumed %Ap -3.4 -4.3 in Limiting EOC Zero Power SLB)

Tilt Allowance  % 3.0 3.0 Safety Parameters Power Level MWt 2754 2754 Maximum Steady State Core Inlet F 550 550 Temperature Minimum Steady State RCS Pressure psia 2200 2200 6 133.9 Reactor Coolant Flow (550 F, 2200 psia) 101b/hr 135.3 Negative Axial Shape Index LC0 I .14 .16 P

extreme assumed at Full Power Maximum CEA Insertion at Full Power  % of Insertion of 25 25 Bank 5 Maximum Initial Linear Heat Rate for KW/ft 16.0 16.0 Transient Other Than LOCA Steady State Linear Heat Rate to Fuel KW/ft 21.0 21.0 Centerline !1elt Assumed in the Safety Analyses CEA Drop Time from Removal of sec 3.1 3.1 Power to Holding Coils to 90%

Insertion Minimum DNBR (CE-1) 1.19 1.195**

  • The effective initial MTC for the SLB event is -2.2X10 ap/*F.
    • Includes a penalty for fuel rod bowing w

i 7.1 ANTICIPATED OPERATIONAL OCCURRENCES FOR WHICH THE RPS ASSURES NO VIOLATION OF LIftITS l The events in this category were analyzed for Calvert Cliffs Unit 1 l- Cycle 5 to determine that Acceptable Limits on DNBR, fuel Centerline temnerature to Melt (CTM), Reactor Coolant System (RCS) upset nressure, and 10CFR100 site boundary dose rate guidelines will not be exceeded. Each of the event writeups.in the section identifies which criterion the event in question t

addresses. Protection against exceeding these liants will continue to be assured by the Reactor Protective System (RPS) Limit'ing Safety System Settings (LSSS) setpoints. The setpoints will be modified (as necessary) to include changes necessitated by the results of the analyses of these events. The methodology used to generate the Limiting Safety System Settings (LSSS) for the TM/LP and ASI RPS trips are consistent with the methods used in the Reference cycle analyses, i

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7.1.1 CEA WITHDRAWAL EVENT The CEA Withdrawal event was reanalyzed for Cycle 5 to determine that the DNBR and CTM design limits are not exceeded.

As stated in CENPD-199-P (Reference 1), the CEA Withdrawal event initiated at rated thermal power is one of the DBEs analyzed to determine a bias factor used in establ'.shing the TM/LP setpoints. This bias factor, along with conservative temperature, pressure, and power' trip input sianals assures that the TM/LP trip prevents the DNBR from dropping below the SAFDL linit (DNBR=1.195 based on CE-1 correlation) for a CEA Withdrawal event. Hence, this event was analyzed for Cycle 5 to generate the bias term input to the TM/LP trip.

The CEA Withdrawal transient may reeuire protection against exceeding both the DNBR and fuel centerline melt (<W/ft) SAFDLs. Depending on the initial conditions and the reactivity insertion rate associated with the CEA withdrawal, either the Variable High Power Level or Thermal Margin / Low Pressure (TM/LP) trip reacts to prevent exceeding the DNBR SAFDL. An approach to the XW/ft limit is terminated by either the Variable High Power Level trip or the Axial Flux Offset trip.

The zero power case was analyzed to demonstrate that SAFDLs are not exceeded.

For the zero power case, a reactor trip, initiated by the Variable Figh Power trip at 40". of rated thermal power is assumed.

The key parameters for thc cases analyzed are reactivity insertion rate due to rod motion, moderator temperature feedback effects and initial axial power distribution. The Resistance Temperature Detector (RTD) response

. time is also important in determining the pressure bias factor.

The range of reactivity insertion rates considered in the analysis is given in Table 7.1.1-1, along with the values of other key parameters used in the analysis of this event.

The maximum reactivity insertion rate assumed for Cycle 5 is 1.6X10-4ac/sec. This reactivity withdrawal rate was calculated by combining the maximum CEA differential worth of 3.2X10-40/ inch and the maximum CEA Withdrawal speed of 30 inches per mirute.

Scram reactivity versus insertion used in the analysis of both zero and full power cases corresponds to a bottom peaked shape characterized by a ASI of +.5. This power distribution maximizes the time required to terminate the decrease in DNBR following a trip.

The CEA Withdrawal transient initiated at rated thermal power results in the maximum pressure bias factor of 70.0 psia. This bias factor accounts for DNB margin degradation from the time a reactor. trip is initiated until minimum DNBR is reached. This pressure bias factor is used in generating TM/LP trip setpoints to prevent the DNB SAFDL from being exceeded.

The zero power case initiated at the limiting conditions for coeration results in a minimum DNBR of 1.27. Also, the analysis shows that the fuel centerline temperatures are eell below those corresponding to the fuel centerline melt SAFDL.

The Sequence of events for the zero power case is presented in Table 7.1.1-2, Figures 7.1.1-1 to 7.1.1,-4 presents the transient behavior of core power,

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4 core average heat flux, the RCS pressure and the RCS temperatures, The analysis of a CEA Withdrawal event presented herein, shows that the DNB and fuel centerline melt SAFDLs will not be exceeded during a CEA f Withdrawal transient.

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TABLE 7.i.1-1 KEY PARAf1ETERS ASSUMED Ill THE CEA WITH0RAWAL AllALYSIS Reference Unit 1 Parameter Units Cycle

  • Cycle 5 Initial Core Power Level (HZP, HFP) ". of 2700 f-twt 0,102- 0, 102 Core Inlet Coolant Temperature 'F 532, 550 532, 550 (HZP, '!FP)

Reactor Coolant System Pressure psia 2200 22 00 Moderator Temperature Coefficient 10-4ao/ F +.5 +.5 Doppler Coefficient Multiplier .85 .85 CEA Worth at Trip - FP 10-24o -5.14 -4.3 CEA Worth at Trip - ZP 10-2ao -3.4 -4.0 Reactivity Insertion Rate X10~4ao/sec 0 to 1.3 0 to 1.6 Holding Coil Delay Time sec 0.5 0.5 CEA Time to 90 Percent Insertion sec 3.1 3.1 (Including Holding Coil Delay)

Resistance Temperature Detector sec 8.0 8.0 Response Time (T)

Rod Group Withdrawal Speed in/ min 30.0 30.0 Maximum CEA Differential Worth X10~4ao/ inch 2.6 3.2 Cycle 4 - last detailed analysis presented (Reference 2).

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i TABLE 7.1.1-2 SEQUENCE OF EVENTS FOR l CEA WITHDRAWAL FROM ZER0 POWER Time (sec) Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled --

Reactivity Insertion 39.0 High Power Trip Signal Generated 40% of 2700 MWt 39.4 Trip Breakers Open --

39.9 CEAs Begin to Drop Into Core --

40.4 Core Power Reaches Maximum 139% of 2700 MWt 41.5 Core Heat Flux Reaches Maximum 65% of 2700 MWt 41.5 Minimum DNBR Occurs 1.27 43.7 Pressurizer Pressure Reaches itaxinun 2371 r.sia 1

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, 7.1.2 Boron Dilution i

No change from original request except that in Table 7.1.2-1 the refueling i mode shutdown margin for Cycle 5 changes from -9.09%ap to -5.0%Ao. This change does not impact the results or the conclusions stated in the original request.

7.1.3 Startup of an Inactive' Reactor Coolant Pump Event No change from original request.

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t 7.1.4 EXCESS LOAD EVENT The Excess Load Event was reanalyzed to determine that the DNBR and CTM design limit are not exceeded during Cycle 5.

The analyses included the effects of manually tripping the RCP's en SIAS due to low pressurizer pressure and the automatic initiation of auxiliary feedwater flow on low steam generator level trip signal.

The High Power level and Thermal Margin / Low Pressure (TM/LP) trips provide primary protection to prevent exceeding the DNBR limit during this event.

Additional protection is provided by other trip signals including high rate of change of power, low steam generator water level, and low steam generator pressure. In this analysis, credit is taken only for the action of the High Power trip in the determination of the minimum transient DNBR.

The approach to the CTM limit is terminated by either the Axial Flux Offset trip, Variable High Power Level trip or the DNB related trip discussed above.

The most limiting load increase events at full. cower and at hot standby conditions, for approach to the DNBR limit of 1.195 (CE-1), are due to the complete opening of the steam dump and bypass valves.

For conservatism in the analyses, auxiliary feedwater flow rate corresponding to 21% of full power main feedwater flow was assumed (i.e.,10.5% of full power main feedwater flow per generator). Also, the addition of the auxiliary feedwater to each steam generator was conservatively assumed to occur 180 seconds after reactor trip. The addition of the auxiliary feedwater flow to both steam generators results in anadditional cooldown of the RCS and a potential for a return-to-power (R-T-P) or criticality arising from reactivity feedback mechanisms.

The Excess Load event at full power was initiated at the conditions g'ven in Table 7.1.4-1. A Moderator Temperature Coefficient of -2.5X10-4ac/F was assumed in this analysis. This MTC, in conjunction with the decreasing coolant inlet temperature, enhances the rate of increase of heat flux at the time of reactor trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions with an uncertainty of 15% was used in the analysis since this FTC causes the least amount of negative reactivity change for mitigating the transient increase in core heat flux. The minimum CEA worth assumed to be available for shutdown at the time of reactor trip for full power operation is 4.3%ao. The analysis conservatively assumed that the worth of baron injected from the safety injection tank is -1.00%a:

per 105 PPit. The pressurizer pressure control system was assumed to be inoperable because this minimizes the RCS pressure during the event and therefore reduces the calculated DNBR. All other control systems were assumed to be in manual mode of operation and have no impact on the results of this event.

The Full Power Excess Load event results in a High Power trip at 7.3 seconds.

The minimum DNBR calculated for the event at the conditions specifi?d in Table 7.1.4-1 is 136 compared to the design limit of 1.195. The maximum local linear heat generation rate for the event is 18. 3 KW/ft.

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For the Excess Load event initiated from HFP conditiont., SIAS is generated  !

at 33.2 seconds at which time the RCP's are manually tripped by the '

operator. The coastdown of the pumps decreases the rate of decay heat removal and therefore keeps the RCS coolant temperatures and pressure at higher values.

Auxiliary feedwater flow is delivered to both stem generators at 187.3 seconds. The feedwater flow causes additional cooldown of the RCS. The decreasing temperatures in combination with a negative MTC inserts positive reactivity which enables the core to approcch criticality. However, the negative ceactivity incerted due to the CEAs and Boron injected via the High Pressure Safety Injection (HPSI) pumps is sufficient to keep l the core subcritical at all times.

S Table 7.1.4-2 presents the sequence of events for an Excess Load event initiated at HFP conditions. Figures 7.l.4-1 to 7.1.4-5 show the 1 NSSS response for. power, heat flux, RCS temperatures, RCS pressure, 3

and steam generator pressure during this event.

The Zero Power Excess Load event was initiated at the conditions given l in Table 7.1.4-3. The MTC and FTC values assumed in the analysis are l the same as for the full power case for the reasons previously given.

The minimum CEA shutdown worth available is conservatively assumed to be -4.0ildo.

i The results of the analysis show that a variable high power trip occurs at 3 6.8 seconds. The minimum DNBR calculated during the event is 2.31 and the peak linear heat generation rate is 14.1 K!!/ft.

. As with the HFP Excess Load event, an SIAS signal on low pressurizer pressure is generated at 77.1 seconds for the zero power excess load event. At 216,8 seconds auxiliary feedwater flow is delivered to both steam generators. The additional positive reactivity due to the cooldown of the RCS is mitigated by the negative reactivity inserted due to CEA's and the boron injected via the HPSI puaps. The core remains subcritica' at all times during an Excess Load event initiated from HZP conditions.

The sequence of events for the zero power case is presented in Table 7.1.4-4. Figures 7.1.4-6 to 7.1.4-10 show the NSSS response for core power, core heat flux, RCS temperature, RCS pressure and steam generator 1 pressure.

For the full and zero power Excess Load events initiated by a full: opening of the steam dump and bypass valves the DNBR and CTM linits are not exceeded. In addition the core remains subcritical even after automatic initiation of the auxiliary feedwater flow and following manual trip of the RCP's

, on SIAS due to low pressurizer pressure. The reactivity transient during I

a HFP and HZP Excess Load event is less limiting than the corresponding-Steam Line Rupture events (See Section 7.3.2).

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TABLE 7.1.4-1 -

KEY PARAMETERS ASSUMED FOR FULL POWER EXCESS LOAD EVENT ANALYSIS Parameter Units Cycle 5 Initial Core Power Level MWt 2754 ,

Core Inlet Temperature *F 550 Reactor Coolant System Pressure psia 2200 6

Core Mass Flow Rate X10 1bm/hr 133.9 Moderator Temperature Coefficient X10-4ap/ F -2.5 CEA Worth Available at Trip %Ap -4.3 Doppler fiultiplier .85 Inverse Boron Worth PPM /%Ao 105 Auxiliary Feedwater Flow Rate Ibm /sec 175.0/S. G.

High Power Level Trip Setpoint  % of Full Power 112

' Low S. G. Water Leval Trip Setpoint ft. 30.9 Axial Power Distribution Scram Calculation ASI +.24 DNBR Calculation + ASI .24 Reference Cycle is FSAR. Full Power Excess Load results were not presented in FSAR, therefore no comgarison is made.

+ The ASI used is conservative with respect to the most negative ASI allowed by the DNB LC0 I

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i TABLE 7.1.4-2 SEQUENCE OF EVENTS FOR THE EXCESS LOAD EVENT AT FULL POWER TO CALCULATE MINIMUM DNBR Time (sec) Event Setpoint or Value 0.0 Complete Opening of Steam Dump and --

Bypass Valves at Full Power 7.3 High Power Trip Signal Generated 112% of full power 7.7 Trip Breakers Open --

8.2 CEA's Begin to Drop Into Core --

8.3 Maximum Pm.er; 114.3 % of full Maximum Lo al Linear Heat power Rate Occurs, KW/ft 18.3 8.7 Minimum DNBR Occurs 1.36 10.6 Low Steam Generator Level Trip Setpoint Reached 30.9 ft 33,2 Safety Injection Actuation signal 1578 psia Initiated; Manual Trip of RCP's 33.5 Pressurizer Empties ~~

54.6 Main Steam Isolation Signal 548 psia 68.2 Rampdown of Main Feedwater Flow Completed Si,of full power main feedwater flow 91.6 Pressurizer Begins to Refill --

134.6 Isolation of Main Feedwater Flow to Both --

Steam Generators 187.3 Auxiliary Feedwater Flow Delivered to Both 175.0 lbm/sec to Steam Generators each steam generator 600.0 Operator Terminates Auxiliary Feedwater __

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TABLE 7.1.4-3 KEY PARAMETERS ASSUMED FOR HOT STAflDBY EXCESS' 1. DAD EVENT ANALYSIS Reference

  • Parameter Units Cycle Cycle 5 Initial Core Power Level MWt 1 1 Core Inlet Temperature *F 532 532 Reactor Coolant Systen Pressure psia 2250 2200 6

Core Mass Flow Rate X10 1bm/hr 137.1 137.1 Moderator Temperature X10-4Ap/'F -2.5 -2.5 Coefficient CEA Worth Available at Trip %Ao -2.4 -4.0 Doppler Multiplier .85 .85 Inverse Boron Worth PPM /%Ap 87 100 Variable High Power Trip  % of full **

40 Setpoint power Low S. G. Water Level Trip ft. 30.01 30.9 Setpoint Auxiliary Feedwater Flow lbm/sec - -175.0/S. G.

Rate Axial Power Distribution Scram Calculations ASI +.53 DriBR Calculations + ASI .42 Reference Cycle is FSAR No credit was assumed for Variable High Power Trip; Reactor trip occurred on Low S. G. level.

+ The ASI used is conservative with res iect to the most negative ASI allowea by the DNB LC0 l

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TABLE 7.1.4-4 SEQUENCE OF EVENTS FOR EXCESS LOAD EVENT AT HOT STANDBY CONDITIONS TO CALCULATE MINItiUM DNBR Setpoint or Value Event 1.ime(sec)_

Steam Dump and Bypass Valves Open to 0.0 Maximum Flow Capacity 40% of full power Variable High Power Trip Signal Generated 3&8 37.2 Trip Breakers Open 37.7 CEA's Begin to Enter Core 40.3% of full 37.8 Core Power Reaches flaximum p wer 37.8 Maximum Local Linear Heat Generation 14.1 Rate Occurs. KU/ft

2. 31 38.3 Minimum DNBR (CE-1)

Pressurizer Empties 73.4 1578 psia 77.1 Safety Injection Actuation Signal Generated; --

Manual Trip of RCS Coolant Punps Main Steam Isolation Signal Generated 548 psia 83.4 Low Steam Generator Water Level ',' rip Setpoint 30.9 ft 89.3 Reached 119.9 Pressurizer Begins to Refill 163.4 Isolation of Main Feedwater Flow to Both Steam Generators Auxiliary Feedwater Flow Delivered to 175.0 lbn/sec 21 6. 8 tc tsch steam Both Steam Generators generator O

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- TABLE 7.14-4 (CONTINUED) 4

-Time (sec) ' Event Setpoint or Value 600.0 Operator Terminates Auxiliary Feedwater --

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7.1.5 LOSS OF LOAD EVENT The Loss of Load event was reanalyzed for Cycle 5 to determine that the transient DNBR does not exceed the'0NBR design limit of 1.195 (CE-1 correlation) and that the RCS pressure upset limit of 2750 psia is not exceeded.

The assumptions used to maximize RCS pressure during the transient are:

a) The event is assumed to result from the sudden closure of the turbine stop valves without a simultaneous reactor trip. This assumption causes the greatest reduction in the rate of heat removal from the reactor coolant system and thus results in the most rapid increase in primary pressure and the closest approach to the RCS pressure upset limit.

b) Tha steam dump and bypass system, the pressurizer spray system, and the power operated pressurizer relief valves are assumed not to be operable. This too maximizes the primary pressure reached during the transient.

The Loss of Load event was initiated at the conditions shown in Table 7.1.5-1. The combination of parameters shown in Table 7.1.5-1 maximizes the calculated peak RCS pressure. As can be inferred from the table, the key parameters for this event are the initial primary and secondary pressures and the moderator and fuel temperature coefficients of reactivity.

The initial core average axial power distribution for this analysis was assumed to be a bottom peaked shape. This distribution is assumed because it ninimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure and heat flux increases. The Moderator Temperature Coefficient (MTC) of +.5X10-4ao/ F was assumed in this analysis. This MTC in conjunction with the increasing coolant temperatures, maximizes the rate of change of heat flux and the pressure at the time of reactor tri p.

A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions was used in the analysis. This FTC causes the least amount of negative reactivity feedback to mitigate the transient increases in both the core heat flux and the pressure. The uncertainty on the FTC used in the analyses is shown in Table 7.1.5-1. The lower limit on initial RCS pressure is used to maximize the rate of change of pressure, and thus peak pressure, following trip.

The Loss of Load event, initiated from the conditions given in Table 7.1.5-1, results.in a high pressurizer pressure trip signal at 8.3 seconds. At 11.5 seconds, the primary pressure reaches its maximum value of 2550.0 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 3.7 seconds. The secondary i

pressure reaches its maximum value of 1050.0 psia at 11.4 seconds after initiation of the event.

The event was.also reanalyzed with the initial conditions listed in Table 7-2 to' detennine that the acceptable DNBR limit is not exceeded. The minimum transient OtJBR calculated for the event is 1.35 as compared to the design limit of 1.195.

4 Table 7.1.5-2 presents the sequence of events for this event. Figures

, 7.1.5-1 to 7.1.5-4 show the transient behavior of power, heat flux, RCS coolant temperatures, and RCS pressure.

The results of-this analysis demonstrates that during a Loss of Load event the peak RCS pressure and the minimum DTISR do not exceed their respective design limits.

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TABLE 7.1.5-1 .

KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD ANALYSIS TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Reference

  • Parameter Units Cycle Cycle 5 Initial Core Power Level MWt 2754 2754 Initial Core Inlet Coolant 'F 552 550 Temperature 0

Core Coolant Flow X10 lbm/hr 133.5 133.9 Initial Reactor Coolant psia 2250 2200 System Pressure Initial Steam Ge>1erator psia 875.0 864.0 Pressure Moderator Temperature X10~4Ao/*F +.5 +.5 Coefficient Doppler Coefficient .85 .85 Multiplier CEA Worth at Trip %Ao -5.14 -4.7 Time to 90% Insertion of sec 2.5 3.1 Scram Rods Reactor Regulating System Operating Mode Manual Manual Steam Dump and Bypass System Operating itode Automatic Inoperative Cycle 2 - last detailed analyses presented (Reference 3 ).

. . =-. .

TABLE 7.1.5-2 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Time (sec) Event Setpoint or Value 0.0 Less of Secondary Load --

3.7; Steam Generator Safety Valves Open 1000 psia 8.3 High Pressurizer Pressure Trip 2422 psia Signal Generated 4

9.7 CEAs Begin to Drop Into Core --

9.8 Pressurizer Safety Valves Open 2500 psia 11.4 Maximum Steam Generator Pressure 1050 psia 11.5 Maximum RCS Pressure 2550 psia

13.4 Pressurizer Safety Valves are Fully 2500 psia Closed f

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7.1.6 LOSS OF FEEDWATER FLOW EVENT No change from original request.

7.1.7 EXCESS HEAT REMOVAL DUE TO FEEDWATER MALFUNCTION No change from original request.

7.1.8 RCS DEPRESSURIZATION EVENT No change from original request.

I 7.2 ANTICIPATED OPERATIONAL OCCURRENCES WHICH ARE DEPENDENT ON INITIAL OVERPOWER MARGIN AND/0R RPS TRIPS FOR PROTECTION AGAINST VIOLATION OF LIMIT,S, The events in this category were analyzed for Calvert Cliffs Unit 1, Cycle 5 to determine the initial margins that must be maintained by the Tech Spec LCO limits such that acceptable DNBR, CTM and upset pressure limits will not be exceeded during any of these events.

The initial margin required to p* event the appropriate limits from being exceeded for any of these events was determined using the initial conditions specified in Table 7-2. For each event conditions were chosen to assure that sufficient initial overpower margin is available at the initiation of the most limiting A00 in this category. The methodology used to generate Limiting Conditions for Operation (LCO) is identical to the methods used in the Reference cycle analyses.

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7.2.1 LOSS OF COOLANT FLOW EVENT .

The Loss of Coolant flow event was reanalyzed for Cycle 5 to determine the minimum initial margin that must be maintained by the Limiting Conditions for Operations (LCOs) such that in conjunction with the RPS (low flow trip), the DNBR limit will not be exceeded.

The methods used to analyze this event are consistent with the methods used in the Reference cycle analyses.

The 4-pump Loss of Coolant Flow produces a rapid approach to the DNBR limit due to the rapid decrease in the core coolant flow. Protection against exceeding the DNB2 lirit for this transient is provided by the initial steady state thermal margin which is maintained by adhering to the Technical Specifications LCOs on DNB and by the response of the RPS which provides an automatic reactor trip on low reactor coolant flow as measured by the steam generator differential pressure transmitters.

The transient is characterized by the flow coastdown curve given in cigure 7.2.1-1. Table 7.2.1-1 lists the key transient parameters used in the present analysis.

Table 7.2.1-2 presents the NSSS and RPS responses during a four pump loss of flow initiated at a 0.0 shape index. The low flow trip setpoint is reached at 1.00 seconds and the scram rods start dropping into the core at 2.0 seconds. A minimum CE-1 DNBR of 1.195 is reached at 3.3 seconds.

Figures 7.2.1-2 to 7.2.1-5 present the core power, heat flux, RCS pressure, and core coolant tenperatures as a function of time.

The event initiated from the Tech Spec LCOs in conjunction with the Low Flow Trip, will ensure that DNBR will not be exceeded.

Y

TABLE 7.2.1-1 XEY PARAt1ETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter Units Reference Cycle

  • Cycle 5 Infital Core Power Level MWt 2754 2754 Initial Core Inlet Coolant *F 550 550 Temperature 0

Initial Core fiass Flow Rate 10 lbm/hr 135.24 133.9 Reactor Coolant System Pressure psia 2200 22 00 Moderator Terrperature Coefficient 10-4ap/F +.5 +.5 Doppler Coefficient Multiplier --

1.00 1.00**

LFT Response Time sec .5 .5 CEA Holding Coil Delay sec 0.5 0.5 CEA Time to 90% Insertion sec 3.1 3.1 (Including Holding Coil Delay)

CEA Worth at Trip (all rods out) 10-2ao -5.72 -5.60 Unrodded Radial Peaking Factor 1.58 1. 62 (FT) 4-Pump RCS Flow Coastdown Figure 7.3-1 Figure 7.2.1-1 Cycle 4 - last detailed analysis presented (Reference 2 ).

Since this is a second order effect and the nost limiting donoler multiplier varies during the transient, a nominal value is used.

l 1

TABLE 7.2.1-2 SEQUENCE OF EVENTS FOR LOSS OF FL0tl Time (sec) Event Setpoint or Value 0.0 Loss of Power to all Four Reactor ----

1 Coolant Pumps 1

1

1.00 Low Flow Trip Signal Generated 93% ofInitial 4-Pumo Flow 1.50 Trip Breakers Open ----

2.00 Shutdown CEAs Begin to Drop Into Core ----

3.3 Minimum CE-1 DNBR 1.195 6.2 Maximum RCS Pressure, psia 2302 4

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7.2.2 LOSS OF ALL NON-EMERCENCY A-C POWER EVENT No change from original request.

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i 7.2.3 FULL LENGTH CEA DROP EVENT  !

The Full Length CEA Drop event was reanalyzed for Cycle 5 to determine the initial thermal margins that must be maintained by the Limiting Conditions for Operation (LCOs) such that the DNBR and fuel centerline melt design limit will not be exceeded.

The methods used to analyze this event are consistent with the methods used in the Reference cycle analyses.

Table 7.2.3-1 lists the key input parameters used for Cycle 5 and compares them to the reference cycle values. Conservative assumptions used in the analysis include:

1. The most negative moderator and fuel temperature coefficients of reactivity (including uncertainties), because these coefficients produce the minimv1 RCS coolant temperature decrease upon return to 100% power level .nd lead to the minimum DNBR.
2. Charging pumps and proportional heater systems are assumed to be inoperable during the transient. This maximizes the pressure drop during the event.
3. All other systems are assumed to be in manual mode of operation and have no impact on this event.

The event is initiated by dropping a full length CEA over a period of 1.0 second. The maximum increases in (integrated and planar) radial peaking factors in either rodded or unrodded planes were used in all axial regions of the core once the power returns to the initial level. Values of 16%

were assumed for these peak increases at full power. The axial power shape in the hot channel is assumed to remain unchanged and hence the increase in the 3-D peak is prcportional to the maximum increase in radial peaking factor of 16P.. Since there is no trip assumed, the peaks will stabilize at these asymptotic values after a few minutes since the secondary side continues to demand 100% power.

Table 7.2.3-2 presents the sequence of events for the Full Length CEA Groo event initiated at the conditions described in Table 7.2.3-1. The transient behavior of key NSSS parameters are presented in Figures 7.2.3-1 to 7.2.3-4.

The transient initiated at the most negative shape index LCO, ( .16) and at the maximum power level allowed by the LCO, results in a minimum CE-1 DNBR of 1.195. A maximum allowable initial linear heat generation rate of 17.9 Kil/ft could exist as an initial condition without exceeding 21.0 )Gl/ft during this transient. This amount of margin is assured by setting the Linear Heat Rate related LC0's based on the more limiting allowable linear heat rate for LOCA.

Consequently, it is concluded that the Full Length CEA Drop event initiated from the Tech Spec LCOs will not exceed the DNBR and centerline to melt design limits.

TABLE 7.2.3-1 KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS Parameter Units Reference Cycle

  • Cycle 5 Initial Core Power Level MWt 2754 27 54 Core Inlet Temperature F 550 5 50 Reactor Coolant Systen Pressure psia 2200 2Z00 6

Core Mass Flow Rate X10 1bm/hr 134.2 1 3 3. 9 Moderator Temperature Coefficient X10-4Ao/*F -2.5 -2.5 Doppler Coefficient Multiplier --

1.15 1.15 Maximam CEA Insertion at Allowed " Insertion of

. 25 25 Power Bank 5 Dropped CEA Worth %Ao unrodded .07 .04 PDIL .04 Most Negative Axial Shape Index I .21 .16 P

Allowed at Full Power (LCO)

Integrated and Planar Radial Unrodded Region 1.16 1.16 Peaking Distortion Factor Bank Inserted 1.16 1.16 (Full Power) Region Cycle 3 - last detailed analysis presented (Reference 4 ).

TABLE 7.2.3-2 SEQUENCE OF EVENTS FOR CEA DROP Time (sec) Event Setpoint Value f

0.0 CEA Begins to Drop ----

1 1.0 CEA Fully Dropped -0.04%Ao 1.1 Core Power Reaches Minimum 94.1% of 2700 Mwt 4.2 Core Heat Flux Reaches Minimum 100.0% of 2700 Mwt

250. Minimum DNBR Reached 1 .1 95 250. Heat Flux Reaches Final Value 102% of 2700 Mwt 300. Core Inlet Temperature Reaches Minimum 54EL5'F 300. RCS Pressure Reaches Minimum 2179 psia e

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FULL LENGTH CEA DROP Figure Calvert cirffs CORE POWER vs TIME tivclear Power Plant 7.2.3-1

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l 7.2.4 A00'S RESULTING FROM THE MALFUNCTION OF ONE STEAM GENERATOR The transients resulting from the malfunction of one ste . generator were analyzed for Cycle 5 to determine the initial margins tha*. must be maintained by the LCO's such that in conjunction with the RPS (Asymmetric Steam Ganerator Protective trip), the DNBR and fuel centerline melt design limits are not exceeded.

The methods used to analyze these events are consistent with those reported in Section 7.2.3 of Reference 1, except that TORC /CE-1 was used instead of COSM0/U-3 to calculate the DNBR. . In addition, the Asymmetric Steam Generator Protective Trip (ASGPT) replaces low steam generator level trip as the primary trip to mitigate this event. A description of this addition to the RPS is described in Appendix A.

The four events which affect a single generator are identified below:

1. Loss of Load to One Steam Generator
2. Excess Load to One Steam Generator
3. Loss of Feedwater to One Steam Generator
4. Excess Feedwater to One Steam Generator Of the four events described above, it has been determined that the Loss of Load to One Steam Generator (LL/lSG) transient is the limiting asymetric event. Hence, only the results of this transient are reported.

The event is initiated by the inadvertent closure of a single main steam l isolation valve. Upon the loss of load to the single steam generator,

' its pressure and temperature increase to the opening pressure of the i secondary safety valves. The intact steam generator " picks up" theThe lost load, which causes its temperature and pressure to decrease.

cold leg asymmetry causes an inlet temperature tilt which results in an azimuthal power tilt, increased PLHGR and a degraded DNBR.

The LL/lSG was initiated at the conditions given in Table 7.2.d-1.

A reactor trip is generated by the Asymmetric Stean Generator Protection Trip at

3. 5 seconds based on high differential pressure between the steam generators.

Table 7.2.4-2 presents the sequence of events for the Loss of Load to One Steam Generator. The transient behavior of key NSSS parameters are presented . in Figures 7.2.4-1 to 7.2.4-5.

A maximum allowable initial linear heat generation rate of 19.0 KW/ft could exist as an initial condition without exceeding 21.0 KW/ft during this transient. This amount of margin is assured by setting the Linear Heat Rate LC0 based on the more limiting allowable linear heat rate for LOCA.

The event initiated from the extremes of the LCO in conjunction with the ASGP trip will not lead to DNBR or centerline fuel temperatures which exceed the DNBR and centerline to melt design limits.

The minimum transient DNBR calculated for the LL/lSG event is 1.39 as compe. red to the minimum acceptable ONBR of 1.195.

4 TABLE 7.2.4 -1 KEY PARAMETERS ASSUMED IN THE ANALYSIS OF LOSS OF LOAD TO ONE STEAM GENERATOR Reference Parameter Units Cycle

  • Cycle 5 Initial Core Power MWt 2700 2754 Initial Core Inlet *F 552 550 Temperature Initial Reactor Coolant psia 2200 22 00 System Pressure

~4 Moderator Temperature 10 ao/ F -2.5 -2.5 Coefficient Doppler Coefficient --

0.85 0.85 Multiplier This event was not analyzed in the FSAR, but was evaluated in CENPD-199-P (Reference 1). Thus Reference 1 parameters are compared with Cycle 5 parameters.

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l TABLE 7.2.4-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD TO ONE STEAM GENERATOR Time (sec) Event Setpoint or Value 0.0 Spurious closure of a single main steam ----

isolation valve 0.0 Steam flow from unaffected steam generator ----

increases to maintain turbine power 2.6 ASGPT* setpoint reached (differential pressure) 175 psid 3.2 Dump and Bypass valves are open ----

3.4 Safety Values Goen nn isolated steam generator 1000 psia i

3.5 Trio breakers open ----

4.0 CEAs begin to insert ----

5.5 Minimum DNBR occurs 1.39

9. 6 Maximum steam generator pressure 105 4 psia ASGPT - Asymmetric Steam Generator Protection Trip I_

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- 7.3~ POSTULATED ACCIDENTS The events in this category were analyzed for Calvert Cliffs Unit 1, Cycle 5 to ensure acceptable consequences. For these transients some amount of fuel failure is acceptable provided the predicted site boundary dose meets 10CFR100 guidelines.

The following sections present the results of the analyses.

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7.3.1 CEA Ejection Event No change from original request except that the total centerline enthalpy of the hottest fuel pellet and fraction of fuel having a centerline enthalpy 2250 cal /gm were in error and have been cor-rected. Table 7.3.1-2 has been revised to reflect these changes.

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TABLE 7.3.1-2 CEA EJECTION EVENT RESULTS Reference Cycle Unit 1 Full Power Unit 1, Cycle 4 Cycle 5 Total Average Enthalpy of Hottest Fuel Pellet 198. 178.

(cal /gm)

Total Centerline Enthalpy of Hottest Fuel 268.

Pellet (cal /ga) 286.

Fraction of Rods that Suffer Clad Damage 0 0 (average Enthalpy 3; 200 cal /gm)

Fraction of Fuel Having a Least Incipient .01 07 Centerline fielting (Centerline Enthalpy > 250 cal /gm) -

Fraction of Fuel Having a Fully ttolten Center- 0 0 line Condition (Centerline Enthalpy 3; 310 cal /gm)

Reference Cycle Unit'1 Zero Power Unit 1, Cycle 4 Cycle 5 Total Average Enthalpy of Hottest Fuel Pellet 177. 145.

(cal /gm)

Total Centerline Enthalpy of Hottest Fuel 177, 221.

Pellet (cal /gm)

Fraction of Rods that Suffer Clad Damage 0 0 (Average Enthalpy 3; 200 cal /gm)

Fraction of Fuel Having a Least Incipient 0 0 Centerline italting (Centerline Enthalpy > 250 cal /gm) ~

Fraction of Fuel Having a Fully !!alten Centerline 0 0 Condition (Centerline Enthalpy > 310 cal /gm) e

7.3.2 STEAM LINE RUPTURE EVENT No change from original request.

7. 3. 3 STEAM GENERATOR TUBE RUPTURE No change from original request.

7.3.4 SEIZED ROTOR EVENT The Seized Rotor event was reanalyzed for Cycle 5 to demonstrate that the RCS upset pressure limit of 2750 psia will not be exceeded and only a small fraction of fuel pins are predicted to fail during this event.

The methods used to analyze this event are consistent with the reference cycle analysis methods, The single reactor coolant pump shaft seizure is postulated to occur as a consequence of a mechanical failure. The single reactor coolant pump shaft seizure results in a rapid reduction in the reactor coolant flow to the three-pump valu e. A reactor trip for the seized rotor event is initiated by a low coolant flow rate as determined by a reduction in the sum of the steam generator hot to cold leg pressure drops. This signal is compared with a setpoint which is a 6nction of the initial number of operating reactor coolant pumps, ror this event a trip will be initiated when, or before, the flow rate drops to 93 percent of initial (4 pump) flow.

The initial conditions for the Seized Rotor event are listed in Table 7.3.4-1.

These conditions are consistent with the initial conditions assumed for the LOF event (See Sectic 7.2.1). Other assumptions on key parameters are also listed in this table.

In Table 7.3.4-2, the NSSS and RPS responses are shown for the seized rotor event initiated from an axial shape index value IP of .16. The pressurizer pressure reached a maximum value of 2292 psia at 3.50 seconds.

Figures 7.3.4-1 through 7.3.4 4 show core power, core average heat flux, RCS pressure, and coolant temperatures during the transier.t.

A conservatively " flat" pin census distribution (a histogram of the number of pins with radial peaks in intervals of 0.01 in radial peak normalized to the maximum peak) is used to determine the number of pins that experience DNB. The rNults show that the rumber of fuel pins predicted to fail is equal to 3.00*.' in comparison to 0.5% for Cycle 4. This is a slight increase over Cycle 4, and remains a small fraction of the total number of fuel pins.

For the case of the loss of coolant flow resulting from a seizure of a reactor coolant pump shaft, a trip on low coolant flow is initiated to limit the predicted fuel failure to only a small fraction of the total number of pins. Based on the low probability of this event, the small number of predicted fuel pin failures is acceptable. In addition, the maximua RCS pressure experienced durir.' the event 'esill be well under the upset pressure limit of 2750 psia.

w

TABLE 7.3.4-1

~

KEY PARAMETERS ASSUMED IN SEIZED ROTOR ANALYSIS Refdrence Unit 1 Parameter Units Cycle

  • Cycle 5 Initial Core Power Level MWt 2754 2754 Core Inlet Coolant Temperature *F 550 550 4 .' ump Core Mass Flow Rate 0 10 lbm/hr 135.2 133.9 3-Pump Core Mass Flow Rate 6 10 1bm/hr 104.4 103, 4 Reactor Coolant System Pressure psia 2200 22 00 Moderator Temperature Coefficient X10-4ap/*F +.5 +.5 Doppler Coefficient Multiplier --

.85 .85

'CEA Worth at Trip (All Rods Out) %Ap -5.7 -5.6 Unrodded Integrated Radial Peaking 1.58 1.62 Factor with Tilt; F[

Axial Shape Index .15 .16

  • ' Cycle 4 - last detailed analyses presented (Reference 2).

l

TABLE 7.3.4-2 4 SEQUENCE OF EVENTS FOR SEIZED ROTOR Time (Sec) Event Setpoint or Value 0.0 Seizure of One Reactor Coolant --

Pump 0.0 Low Coolant Flow Signal 93% nf Initial 4-Pum; Generated Flow 0.50 Trip Breakers Open --

1.00 CEAs Begin Dropping into Core --

3.50 Maximum RCS Pressure, psia 2292 I

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8. ECCS ANALYSIS No change from original request.
9. TECHNICAL SPECIFICATI0hlSL (No change from original request except where noted)

In this section all changes that must be made to the Technical Specifications are provided in order to make the Technical Specifications valid for operation of Cycle 5.

Each page from the Technical. Specifications which must be modified is shown with the modification included.

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