ML20038A848

Application for Amend to Licenses DPR-53 & 69,consisting of Portions of Cycle 6 Reload Application Entitled Phase I 6.0 Thermal Hydraulic Design

ML20038A848
Person / Time
Site:	Calvert Cliffs
Issue date:	11/19/1981
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	Clark R Office of Nuclear Reactor Regulation
Shared Package
ML20038A849	List: ML20038A848 ML20038A860
References
NUDOCS 8111240251
Download: ML20038A848 (52)

Text

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BALTI M ORE GAS AND 2

ELECTRIC CHARLES CENTER P. O. BOX 1475. BALTIMORE, MARYLAND 21203 i

ARTHUR C. LUNDVALL. JR.

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November 19, 1981 4

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Office of Nuclear Reactor Regulation tg Washington, D. C.

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ATTENTION:

Mr. R. A. Clark, Chief j

l Operating Reactors Branch #3 Division of Licensing Subj ect: Calvert Cliffs Nuclear Power Plant Units No. 1 and 2 Docket Nos. 50-317 and 50-318 Phase I Cycle 6 Reload Application for Amendment to Operating License Reference (A):

A. E. Lundvall, Jr. to R. A. Clark letter, dated 9/22/81, Fifth Cycle License Application Gentlemen:

1 At an October 15, 1981 meeting with NRC staff in Bethesda we agreed to make an early submittal of a portion of the Cycle 6 reload application.

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That portion is titled ' Phase l' and is attached hereto. Specifically, Phase I consists of the following sections of a standard reload application:

i

-- Chapter 6.0 Thermal Hydraulics Design

-- Chapter 7.0 Transient Analyses 7.1.4 Excess Load Event i

7.1.5 Loss of Load Event 7.2.3 Full Length CEA Drop Event j

7.2.4 A00's Resulting from the Malfunction of O(

One Steam Generator

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-- Chapter 9.0 Technical Specifications l

(1111240251 8111'19 I

PDR ADOCK 05000317 j

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U. S. Nuclear Regulatory Commission - 2 November 19, 1981 Attention:

Mr. R. A. Clark (Nine (9) of the anticipated total of twelve (12) modifications to Technical Specifications are included in Phase I.)

The sections of Phase I are not significantly different from those submitted in Reference (A). Phase II of the Cycle 6 application will include Phase I as well as the rest of the sections which constitute a standard reload application. Phase II will be submitted on or about February 15, 1982.

Very truly yours, BALTIMORE GAS AND ELECTRIC COMPANY b

J-mtw }.n/1. ~

A. E. fundvall, Jr.

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Vice President-Supply Attachment Copies to:

J. A. Biddison, Esquire (w/o encl.)

G. F. Trowbridge, Esquire (w/o encl.)

Mr. D. H. Jaffe - NRC Mr. P. W. Kruse - CE

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ATTACllMENT PHASE I 6.0 THERMAL HYDRAULIC DESIGN f

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w 6.0 THERMAL HYDRAULIC DESIGN 6.1 DNBR Analysis Steady state DN5R analyses of Cycle 6 at the rated power level of 2700 MWt

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have been performed using the TORC computer code described in Reference 1, the CE-1 critical heat flux correlation described in Reference 2, and the simplified modeling methods described in Reference 3.

A variant of TORC called CETOP, optimized for simplified modeling applications, was used in this cycle to develop the " design thermal margin model" described generically in Reference 5

Details of CETOP are similar discussion of CETOP methodology we2 discussed ir. Reference 4; E submitted on the Arkansas Nuclear One Unit 2 (ANO-2) docket in Reference CETOP was approved for use on AND-2 in Reference 6.

In general, this 5.code differs from earlier versions of TORC only in that enthalpy transport coefficients are used to improve modeling of coolant conditions in the vicinity of the hot subchannel and in that more rapid equation-solving Direct comparisons show that CETOP models tend to be routines are used.

slightly more conservative than TORC cesign models in computing minimum DrSR for limiting cases.

(Note that application of the methods of Reference 3 assures that design models set up with either TORC or CETOP CETOP is are always conservative relative to detailed TORC analyses.)

used only because it reduces computer costs significantly; no margin gain is expected or taken credit for.

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 note that the calculational factors flux factor, engineering factor on hot channel heat (engineering heat rod pitch and clad diameter factor) listed in Table 6-1 have been

input, combined statistically with other uncertainty factors at the 95/95 confidence / probability level (Reference 7) to define a new design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 7.

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 8 which concluded that the wear problem and the sleeving repair do not adversely affect DNBR margin.

6.2 Effects of Fuel Rod Bowing on DNSR Margin The fuel rod bowing effects on DNB margin for Calvert Cliffs-1 Cycle 6 have been evaluated according to the guidelines set forth in Reference 9.

A total of 137 fuel assemblies will exceed the NRC-specified DNS penalty threshold burnup of 24,000 MWD /T during Cycle 6, as established by the guidelines in Reference 9.

At the end of Cycle 6, the maximum burnup Based upon an attained by any of these assemblies will be 42,800 MWD /T.

extrapolation of the formula contained in Reference 6, the corresponding DNB penalty for 42,800 MWD /T has been determined to be 6.3 percent.

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4 l

An examination of power distributions for Cycle 6 shows that DNS margin exists for assemolies exceeding 24,000 MWD /T relative to the DNB limitsTh the Reference 9 reduction penalty of 6.3 percent imposed upon fuel

{l estaolishec by other assemolles in the Core.

Therefore, no power penalty asemblies exceeding 24,000 MWD /T in Cycle 6.

I for fuel rod bowing is required in Cycle 6.

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t TABLE 6-1 Calvert Cliffs Unit 1 Thermal-Hydraulic Parameters at Full Power Reference General Characteristics _

M Cycle 5*

Cycle 6**

[

MWT 2700 2700 106 Btu /hr 9215 9215 i j-Total Heat Output (core only)

.975

.975 Fraction of Heat Generated in Fuel Rod Primary System Pressure psia 2250 2250 Nominal psia 220]

Minimum in steady state psia 2300 Maximum in steady state 548

• l 550 Inlet Temperature 381,600 370,000 gpm Total Reactor Coolant Flow 106 lb/hr 139.0 143.8 (steady state) 133.9 138.5 106 lb/hr Coolant Flow Through Core ft 0.044 0.044 Hydraulic Diameter (nominal channel) 6 2

2.51 2.61 10 lb/hr-ft Average Mass Velocity 10.4 11.1 psi Pressure Drop Across Core (minimum steady state flow irreversible ap over entire fuel assembly) 34.4 32.4 psi Total Pressure Orop Across Ve.;sel (based on nominal dimensions and mir.imum steady state flow) 186.435***

186,435 ***

Core Average h: ; Flux (accounts Btu /hr-ft for above fraction of heat generated in fuel rod and axial densification factor) 2 48,192***

48,192 ***

ft Total Heat Transfer Area (Accounts for axial densification factor) 5930 5765 Film Coefficient at Average Conditions Btu /hr-ft

• F 32 31

• F Average Film Temperature Difference 6.23***

6.23***

kw/ft werage Linear Heat Rate of Undensified Fuel Rod (accounts for above fraction of heat generated in fuel rod) 66.5 Btu /lb

'68.8 Average Core Enthalpy Rise 657 F

657 Maximum Clad Surface Temperature

i TABLE 6-1 (cont'd)

Reference Calculational Factors Cycle 5 Cycle 6 Engineering Heat Flux Factor 1.03 1.03 * **

• Engineering Factor on Hot 1.02 ~

1.02****

Channel Heat Input Rod Pitch and Clad Diameter 1.065

1. 06 5 ** *

• Factor Fuel Densification Factor (axial) 1.01 1.01 NOTES

• Design inlet temperature and nominal primary system pressure were used to calculate these parameters.

• • Due to the statistical combination of uncertainties described in References 7,10 and 11, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.

• • • Based on a generic value of 1100 shims.

• • • • These factors have been combined statistically with other uncertainty factors at 95/95 confidence / probability level (Reference 7) to define a new design limit on CE-1 midmum DNBR when iterating on power as discussed in Reference 7.

t References (Chapter 6) 1.

CENPD-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," July 1975 4

2.

CENPD-162-P-A (Proprietary) and CENPD-162-A (Nonproprietary), " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1, Uniform Axial Power Distribution," April 1975 4

3.

CENPD-206-P, " TORC Code, Verification and Simplified Modeling Methods,"

January 1977 4.

Letter, A. E. Lundvall, Jr. to R. A. Clark, " Response to Questions on SCU, CEN-124(B)," June 2,' 1981 5.

Letter, D. C. Trimble (AP&L) to Director, NRR, "CETOP-D Code Structure and Modeling Methods, Response to First Round Questions on the Statistical Combination of Uncertainties Program (CEN-139(A)-P)", July 15,1981 a

6.

Final Safety Evaluation Report Supporting Facility Operating License Amendment No. 26 on Docket No. 50-368 and Operation of ANO-2 During Cycle 2, July 21,1981 7.

CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 2," January 1980 8.

CEN-83(B)-P, "Calvert Cliffs Unit 1 Reactor Operation With Modified CEA Guide Tubes," February 8,1978, and letter, A. E. Lundvall, Jr. to V. Stello, Jr.,

" Reactor Operation With Modified CEA Guide Tubes," February 17,1978 9.

Letter, D. F. Ross and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller (NRC), " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculation for Light Water Reactors," February 16, 1977 i

10.

CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 1," January 1980 11.

CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 3," March 1980 i

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7.0 DESIGN BASIS EVENTS 7.1.4 EXCESS LOAD EVENT 7.1.5 LOSS OF LOAD EVENT 7.2.3 FULL LENGTH CEA DROP EVENT A00'S RESULTING FROM THE MALFUNCTION OF ONE STEAM GE 7.2.4 O

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7.1.4 EXCESS LOAD EVENT Excess Load Event was reanalyzed to determine that the ONBR and CTM The design limits are not exceeded during Cycle 6.

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

it-The High Power level and Thermal Margin / Low Pressure (TM/LP) trips provMe 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 In this analysis, credit is taken only for the action generator pressure.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 power and at hot standby conditions, for approach to the DNBR limit of 1.23 (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 generater). Also, the addition of the auxiliary feedwater to each steam generator was conservatively assumed to occur The addition of the auxiliary feecwater 180 seconds after reactor trip.

flow to both steam generators results in anadoitional 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 given A Moderator Temperature Coefficie'nt of -2.5X10-4c/F was in Table 7.1.4-1.

assumed in this analysis. This MTC, in conjunction with the decreasing coolant inlet temperature, enhances the rate of increase of heat flux at A Fuel Temperature Coefficient (FTC) corresponding the time of reactor trip.

to beginning of cycle conditions with an uncertainty of 15% was used in analysis since this FTC causes the least amount of negative reactivity the The change for mitigating the transient increase in core heat flux.

minimum CEA worth assumed to be available for shutdown at the time of reactor trip for full power operation is 4.3%oo. The analysis conservatively assume. that the worth of boron injected from the safety injection tank is -1.00%Ao The pressurizer pressure control system was assumed to be per 105 PPM.

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 & Hip Power trip at 7.2 seconds.

The minimum DNBR calculated for the event at the conditions specified in Table 7.1.4-1 is 1.48 compared to the cesign limit of 1.23.

The r.aximum local linear heat generation rate for the event is 18.4 KW/ft.

.' For the Excess Load event initiated from HFP conditions, SIAS is generated at 34.3 seconds at which t.*' the RCP's are manually tripped by the

.The coastdown of the pumps decreases the rate of decay heat operator.

renoval and therefore keeps the RCS coolant temperatures and pressure at higher values.

Auxiliary feedwater flow is delivered to both steam generators at 187.2 I

The feedwater flow causes additional cooldown of the RCS.

The seconds.

decreasing temperatures in combination with a negative MTC inserts positive

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reactivity whicn enables the core to approach criticality. The negative reactivity inserted due to the CEAs and Boron injected via the High Pressure Sa.fety Injection (HPSI) pumps however is sufficient to maintain the core subtritical at all times.

Table 7.1.4-2 presents the sequence of events for an Excess Load event initiated at HFP conditions.

Figures 7.1.4-1 to 7.1.4-5 show the NSSS response for power,-heat flux, RCS temperatures, RCS pressure, and steam generator pressure during this event.

The Zero Power Excess Load event was initiated at the conditions given The MTC and FTC values assumed in the analysis are in Table 7.1.4-3.

the same as for the full power case for the reasons previously given.

The minimum CEA shutdown worth available is conservatively assumed to be -4.0 2c.

The results of the analysis show that a variable high power trip occurs l

at 35.g seconds. The minimum DNBR calculated during the event is 2.92 and the peak linear heat generation rate is 14.4 KL'/ft.

As with the HFP Excess Load event, an SIAS signal on low pressurizer is generated at 76.6 seconds for the zero power excess load pressure At 215.9 seconds auxiliary feedwater flow is delivered to bcth event.

The additional positive reactivity due to the cooldown stean generators.

of the RCS is mitigated by the negative reactivity inserted due to CEA's and the boron injected via the liPSI pumps. The core remains subtritical 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 Figures 7.1.4-6 to 7.1.4-10 show the NSSS response for core 7.1.4-4 power, core beat flux, RCS temperature, RCS pressure and steam generator 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 In addition the core remains subcritical even after automatic exceeded.

initiation of the auxiliary feedwater flow and following manual trip of the RCP's on SIAS due to lcw pressurizer pressure..The reactivity transient during a HFP and HZP Excess Load event is less limiting than' the corresponding Steam Line Rupture events (See Section 7.3.2).

TABLE 7.1.4-1 1

KEY PARAMETERS ASSUMED FOR FULL POWER EXCESS LOAD EVENT ANALYSIS Reference Units Cycle Cycle 6 Parameter

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2700+

i Initial Core Power Level MWt 2754 548,

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• F 550

ore Inlet Temperature 2225 +

2200 Reactor Coolant System Pressure psia X10 lbm/hr 133.9 138.5+

0 Core Mass Flow Rate Moderator Temperature Coefficient X10 ap/*F

-2.5

-2.5

-4

-4.3

-4.3

%ao CEA Worth Available at Trip

.85

.85 Doppler Multiplier PPM /%do 105 105 Inverse Boron Worth Auxiliary Feedwater Flow Rate ibm /sec 175.0/5.G.

175.0/S. G.

% of Full Power 112 110 High Power Level T' rip Setpoint Low S. G. Water Levr Trip Setpoint ft.

30.9 30.9 Reference cycle is Cycle 5, Reference 2.

+For DNBR calculations, effects of uncertainties on these parame%rs were combined statistically (see Reference 1) l

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TABLE 7.1.4-2 SEQUENCE OF EVENTS FOR THE EXCESS LOAD EVENT AT FULL POWER TO CALCULATE MINIMUM DNBR Setcoint or Value Time (sec)

Event 0.0 Complete Opening of Steam Dump and Bypass Valves at Full Power 110% of full power 7.2 High Power Trip Signal Generated 7.6 Trip Breakers Open 8.1 CEA's Begin to Drop Into Core 113.2% of. full powcr 8.6 Maximum Power; Maximum Local Linear Heat 18.4 Rate Occurs, KW/ft 1.48 Minimum DNER Occurs 9.0 Low Steam Generator Level Trip Setpoint Reached 30.9 ft 10.6 34.1 Pressurizer Empties Safety Injection Actuation Signal Initiated; 1578 psia 34.3 Manual Trip of RCP's

-52.5 tiain Steam Isolation Signal 548 psia 68.1 Rampdown of Main Feedwater Flow Completed 5% of full power main feedwater flew 96.5 Pressurizer Begins to Refill 132.5 Isolation of Main Feedwater Flow to Both Steam Generators 175.0 lbm/sec to.

Auxiliary Feedwater Flow Delivered to Both 187.2 each steam Steam Generators generator Operator Terminates Auxiliary Feedwater

-600.0 Flow to Both Steam Generators

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TABLE 7.1.4-3 KEY PARAMETERS ASSUMED FOR HOT STANDBY EXCESS LDAD EVENT An'ALY Reference

• Units Cycle Cycle 6 i

Parameter 1+

ll Initial Core Power level MWt i

532+

• F 532 Core Inlet Temperature

- 2225+

Reactor Coolant System Pressure psia 2200 Core Mass Flow Rate X10 1bm/hr 137.1 141.35+

6 X10'4ao/*F

-2.5

-2.5 Moderator Temperature Coefficient

-4.0

-4.Q CEA Worth Available at Trip

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.85

.85 Doppler Multiplier 100 Inverse Boron Worth PPM /%ao 100 40 Variable High Power Trip

% of full 40 power Setpoint 3".9 Low S. G. Water Level Trip ft.

30.9 Setpoint Auxiliary Feedwater Flow lbm/sec 175.0/S.G.

175.0/5. G.

Rate Reference Cycle is Cycle 5 in Reference 2.

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

+

statistically (see Reference 1).

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TABLE 7.1.4 4 SEQUENCE OF EVENTS FOR EXCESS LOAD EVENT AT HOT STANDBY CONDITIONS TO CALCULATE MINIMUM CNBR Setooint er Value f;I Time (sec)

Event i' !

I Steam Dump and Bypass Valves Open to 0.0 Maximum Flow Capacity 40% of full powit Variable High Power Trip Signal Generated 35.9 Trip Breakers Open 36.3 40.4% of full Core Power Reaches flaxinum 36.9 2.92 Minimum DNBR (CE-1) 37.6 Pressuri:er Empties 72.3 1578 psia Safety Injection Actuation Signal Generated; 76.6 fianual Trip of RCS Coolant Pumps 548 psia Main Steam Isolation Signal Generated 82.6 30.9 ft Low Steam Generator Water Level Trip Setpoint 88.7 Reached Pressurizer Begins to Refill 106.8 Isolation of Main Feedwater Flow to Both 162.6 Steam Generators 175.0 lbn/sec Auxiliary Feedwater Flow Delivered to to each steam 21 5.9 Both Steam Generators Y

generator 9

TABLE 7.14-4 (CONTINUED)

Set::cint or Value Time (sec)

Event 600.0 Ocerator Terminates Auxiliary Feedwater Flow to Both Steam Generators I i i

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1, References for Section 7 (Hon-LOCA Transien_t Analysis)

" Statistical Conbination of Uncertainties tiethodology; Part 1:

C-E Calculated Local Power Density and Thernal l!argin/Lew Pressure i

1.

f:

LSSS for Calvert Cliffs Units I and II," CEH-124(B)-P, Decenber,1979.

2 i

Part 2:

I :

" Statistical Combination of Uncertainties ilethodology:

Combination of System Paraneter Uncertainties in Thernal 11argin Analyses for Calvert Cliffs Units I and II," CEH-124(B)-P, January,1980.

" Statistical Combination of Uncertainties liethodology; Part 3:

C-E Calculated Local Power Density and Departure from Huc1cate Boiling Limiting Conditions for Operation for Calvert Cliffs Units 1 and II," CEH-124(B)-P,11 arch,1980.

A. E. Lundvall to R. A. Clark, Calvert Cliffs fluclear Power Plant - I !

Docket No. 50-317, "Anendnent to Operating License DPR-53 Supplement 2.

Sth Cycle License Application," November 4,1980.

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7.1.5 LOSS OF LOAD EVENT The loss of Load event was reanalyzed for Cycle 6 to determine that the transient DNBR does not exceed the new design limit and that the RCS a

pressure upset limit of 2750 psia is not exceeded, I

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

The event is assumed to result from the sudden closure of the turbine j'

a)

This assumption stop valves without a simultaneous reactor trip.

causes the greatest reduction in the rate of heat removal from the reactor ccolant system and thus results in the most rapid increase in primary pressure and the closest approach to the RCS pressure upset limit.

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

the transient.

The Loss of Load event was initiated at the conditions shown in Table The combination of parameters shown in Table 7.1.5-1 maximizes 7.1.5-1.

As can be inferred from the table, the calculated peak RCS pressure.

the key parameters for this event are the initial primary and secondary pressures and the noderator and fuel temperature coefficients of reactivity.

The initial core average axial power distribution for this analysis was This distribution is assumed because assumed to be a bottom peaked shape.

it minimizes the negative reactivity inserted during the initial portion of the scram following a reactor trip and maximizes the time required to The Moderator Temperature mitigate the pressure and heat flux increases. Coefficient (MTC) of +.5 x 10 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 A Fuel Temperature Coefficient (FTC) corresponding to beginning trip.

This FTC causes the least of cycle conditions was used in the analysis.

amount of negative reactivity feedback to initigate the transient increases The uncertainty on the in both the core heat flux and the pressure.

The lower limit on FTC used in the analyses is shown in Table 7.1.5-1.

initial RCS pressure is used to maximize the rate of change of pressure, end 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 secandary pressure is limited.by the opening of The secondary the main steam safety vahes, which open at 3.7 seconds.

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

Figures Table 7.l.5-2 pre;ents the sequence of events for this event.show the tra 7.1.5-1 to 7.1.5-4 coolant temperatures, and RCS pressure.

l

t The event was also reanalyzed with the initial conditions listtd in Table 7.1.5-3 to determine that the acceptable DNBR limit is not The minimum transient DNBR calculated for the event is 1.38 exceeded.

as compared to the design limit of le23.

f I

The results of this analysis demonstrates that during a Loss of Load event the peak RCS pressure and the minimum DNBR do not exceed their i{

respective design limits.

O e

i I

I l

l

TAflLE 7.1.5-1 KEY PARA' ETERS ASSUt:ED IN THE LOSS OF LOAD ANALYSIS TO liAXIlil2E CALCULATED RCS PEAR, PRESSURE I

Reference

• Units _

Cycle Cycle 6 I

Parameter 2754 2754 MWt Initial Core rewer Level 550

'F 550 Initial Core, Inlet Coolant Temperature 133.9 0

X10 lbm/hr 133.9 Care Coolant Flow 2200 Initial P.eacter Ccolant psia 2200 System Pressurc 864.0 864.0 Initial Steam Generator psia Pressure

+.5 X10-4ap/*F

+.5 Moderator Temperature Ccefficient

.85

.85 Doppler Ccefficient Multiplier

-4<7

%Ap

-4.7 CEA 1.' orth at Trip 3.1 3.1 sec Time to 90!; Ir.sertion of Scram Rods Manual Manual Reactor Regulating System Operating Mode Inoperative Steam Dump and Bypass System Operating flode Inoperative Cycle 5 (Reference 2)

O

/

s

~

TABLE _'.1.5-2 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE I

Time (sec)

Event Setooin_t or Value j

0.'O Loss of Secondary Load 3.7; Steam Generator Safety Valves Open 1000 psia 8.3 High Pressurizer Pressure Trip 2422 psia Signal Generated 9.7 CEAs Begin to Drop Into Core S.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

TAJLE 7.1.5-3_

KEY PARN:ETERS ASSU :ED It! THE LOSS OF LOAD ANALYSIS TO CALCULATE TRANSIENT MINIMUM DNBR Reference

• Units _

Cycle Cycle 6 Param ter_

2700 2754 MWt Initial Core Tcwer Level

!f

'F 548 initial Core

• Inlet Coolant 550 Temperature X10 'lbm/hr 133.9 138.5 I

Core Coolant Flow Initial Reacter Ccolant psia 2225 2200 System Pressuro Initial Steam Generator psia 864.0 864.0 Pressure 1.75 Integrated Radial Peaking 1.71 Factors, Ft (Bank 5 inserted 25%)

+.5 XID-Oop/*F

+.5 Moderator Temperature Coefficient

.85

.85 Doppler Ccefficient Multiplier

-4=7

%AP

-4.7 CEA Uorth at Trip 3.1

~

3.1 sec Time to 90" Insertion of Scram Rods Manual Manual' Operating Mode Reactor Regulating System Inoperative Steam Ducp and Bypass System Operating liede.

Inoperative 0 ' Cycle 5 (Reference 2)

" Effects of uncertainties on these parameters were accounted for statistica l

(SeeReference1) l l

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CORE POWER vs TIME ~

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REACTOR COOLANT SYSTEM TEMPERATURES vs TIME 7.1.5-4

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7.2.' 3 FULL LENGTH CEA DROP EVENT _

The Full Length CEA Droo evcnt was reanalyzed for Cycle 6 to determine the initial thermal margins that must be maintained by the melt design limit will not be exceeded.

The methods used to analyze this event are consistent with those discussed in Reference 1 except CETOP/CE-1 was used instead of TORC /CE-1 to calculate DN Table 7.2.3-1 lists the key input parameters used for Cycle 6 and compares a

Conservative assumptions used in the them to the reference cycle values.

ar,alysis include:

The most negative moderator and fuel temperature coefficients of reactivity (including uncertainties), because these coefficients 1.

produce the minimum RCS coolant temperature decrease upon return to 100% power level and lead to the minimun DNBR.

Charging pumps and proportional heater systems are assumed to be 2.

inoperable during the transient.

during the event.

All other systems are assumed to be in manual mode of operation and

.3.

have no impact on this event.

The event is initiated by dropping a full length CEA over a period of The maximum increases in (integrated and planar) radial peaking factors in either rodded or unrodded planes were used in all axial regions 1.0 second.

Values of 16%

of the core once the power returns to the initial level.The axial power shape were assumed for these peak increases at full power.

in the hot channel is assumed to remain unchanged and hence the increase in the 3-D peak is proportional to the maximum increase in radial peaking Since there is no trip assumed, the peaks will stabilize factor of 16%.

at these asynptotic 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 D The event initiated at the conditions described in Table 7.2 to 7.2.3-4.

The transient initiated at the most negative shape index LCO (.15) and at the maximum power level allowed by the LCO, remits in a minimum A maximum allowable initial linear heat generation rate of 18.2 KW/ft could exist as an initial condition without exceeding CE-1 DNBR of 1.23.

This amount of margin is assured by 21.3 KU/ft during this transient.

setting the Linear Heat Rate related LCO's based on the more limiting allowable linear heat rate for LOCA.

Consequently, it is concluded that the Full Length CE to nelt design limits.

TABLE 7.2.3-1_

KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYS Units Reference Cycle

• Cycle 6 Parameter _

l 2700 +

2754 MWt Initial Core Power Level 548+

'F 550 Core Inlet Temperature 2225+'

2200 c. Reactor Coolant Systen Pressure psia 138.'5 +

6 X101bm/hr 133.9 Core Mass Flow Rate

-2.5 X10'4ao/*F

-2.5 Moderator Temperature Coefficient 1.15 1.15 Doppler Coefficient Multiplier 25

% Insertion of 25 Maximum CEA Insertion at Allowed Bank 5 4

Power

%oo unrodded

.04

.04 Dropped CEA Worth PDIL

.04

.04 '

Most Negative Axial Shape Index

.16

.15+

Allowed at Full Power (LCO)

Unrodded Region 1.16 1.16 Integrated and Planar Radial 1.16 Bank Inserted 1.16 Peaking Distortion Factor (Full Power)

Region

• Cycle 5 (Reference 2)

For DNBR calculations, effects of uncertainties on these pararaeters were com

+

statistically. (See Reference 1) 9' t

i

~ - -

TABLE 7.2.3-2 SEQUENCE OF EVENTS FOR CEA DROP 1:

Setooint Value_

Event _

Time (sec).

i

.0 CEA Begins to Drop

-0.04%Ap CEA Fully Dropped 1.0 92.2%

Core Power Reaches Minimum 1.1 98.1%

Core Heat Flux Reaches Minimum 4.2 100%

Heat Flux Reaches Final Value 300.

546.5'F Core Inlet Temperature Reaches Minimum 300.

2204.3 psia RCS Pressure Reeches Minimum 300.

1.23 Minimum DNBR Reac!ied 300.

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CORE POWER vs TIME 7.2.3-1 I

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GAS & ELECTRIC CO' CORE AVERAGE HEAT FLUX vs TIME 7.2.3-2I Ccivert Ciiri, Nucle:r ?ower Plcr.t

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a A00'S RESULTING FROM THE MALFUNCTION OF ONE STEAM GENERAT

.7.2.4 The transients resulting from the malfunction of one steam generator were analyzed for Cycle 6 to determine the initial margins that must be maintained by the LCO's such that in conjunction with the RPS (Asymmetric Steam Generator Protective trip), the DNSR and fuel centerline melt design limits are not exceeded.

events are consistent with those reported The methods used to analyze these in the reference cycle, except that CETOP/CE-1 was used instead of TORC /CE-l to calculate the DNBR.

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 Loss of Feedwater to One Steam Generator 3.

Excess Feedwater to One Steam Generator 4.

Of the four events described above, it has been determined that the loss of Load to One Steam Generator (LL/lSG) transient is the limitin asymetric event.

The event is initiated by the inadvertent closure of a single main steam Upon the loss of load to the single steam generator, its pressure and temperature increase to the opening pressure of the isolation valve.

The intact steam generator " picks up" the secondary safety valves.

The lost load, which causes its temperature and pressure to decrease.sn inl cold leg asymmetry cause:

azimuthal power tilt, increased PLHGR and a degraded DNBR.

The LL/lSG was initiated at the conditions given in Table 7.2.4-1.

A reactor trip is generated by the Asymmetric Steam Generator Protection 2.6 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 l

The transient behavior of key NSSS parameters are One Steam Generator.

presented in Figures 7.2.4-1 to 7.2.4-5.

A maximum allowable initial linear heat generation rate of 19.3 KW/ft could exist as an initial condition without exceeding 21.3 KW/ft during 1

This amount of margin is assured by setting the this transient.

Linear Heat Rate LCO 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 DNSR and centerline to melt design limits.

The minimum transient DNBR calculated for the LL/lSG event is 1.43 compared to the minimum acceptable DNBR of 1.23.

TABLE 7.2.4 -1_

KEY PARAMETERS ASSUMED IN THE ANALYSIS OF LOSS OF LOAD'TO ONE STEAM GENERA

[

Reference Cycle J5 Cycle

• _

~

Units _

Parameter _

I 2700 +

MWt 2754 Initial Core Power 548+

550

'F Initial Core Inlet Temperature 2225+

2200 Psia Initial Reactor Coolant System Pressure

-2.5

~4 10 ao/*F

-2.5 Moderator Temperature Coefficient 0.85 0.85 Doppler Coefficient Multiplier f

l

• Cycle 5 (Reference 2)

For DNBR calculations, effects of uncertainties on these parameters were 1

+

combined statistically.

(See Reference 1)

I l

l l

TABLE 7.2e4-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD TO ONE STEAM GENERATOR Setpoint or Value f

Time (secl Event _

e Spurious closure of a single main steam 0.0 isolation valve Steam flow from unaffected steam generator 0.0 increases to maintain turbine power setpoint reached (differential pressure) 175 psid 2.6 ASGPT*

Dump and Bypass valves are open 3.2 3.5 Trip breakers open CEAs begin to insert 4.0 1000 psia Safety valves open on isolated steam generator 4.0 1,43 Minimum DNBR occurs 5.5 1050 psia Maximum steam generator pressure 10.1 ASGPT - Asymmetric Steam Generator Protection Trip

/

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100 s

E 80 8

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0 25 50 75 100 125 150 0

TIME, SECONDS BALTIMORE LOSS OF LOADl1 STEAM GENERATOR EVENT Figure GAS & ELECTRIC CO.

CORE POWER vs TIME 7.2 A 1

~

Colvert Citris Nuclear ?:wer Plent

1 120 i

100 E

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g 5::1 8*

x' 60 3

u W6 x

40 g

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0 25 50 75 100 125 150 0

TIME, SECONDS f

i LOSS OF LOADIl STEAM GENERATOR Ev'ENT Figure oAsI#'EtEcTEic co.

CORE AVERAGE HEAT FLUX vs TIME

~

l coivert citifs 1

Nuclear Power P!:nt

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t-1050 l-

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ISOLATED STEAM GENERATOR

{

vt 950 e

E

$u 900 e

f2 UNIS0 LATED STEAM GENERATOR b

850 M

800 I

0 25 50 75 100 125 150 750 TIME, SECONDS Reure LOSS OF LOADll STEAM GENERATOR EVENT

^"

STEAM GENERATOR PRESSURE vs TIME 7,2 A-3 GA5 E E T IC CO.

cotvert ctins Nuclect Power Plant

O 2300' t

2200 -

2100 G

c.

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Pn2000 O

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e 1900 1800 i

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BALMORE REACTOR COOLANT SYSTEM PRESSURE vs 7.2A-4 j:

GAS & ELECTRIC CO' Carvert Cirrrs Nuclect Power Ple.t

I 615 t

600 -

T0lH 585 vi E

i2<

5 570 AV W

G

" 555 540 0

25 50 75 100 125 150 525 TIME, SECONDS LOSS OF LOADll STEAM GENERATOR EVENT rigore sALTe. sore As IIIe $;h, REACTOR COOLANT SYSTEM TEMPERATURE vs T 7,2 A-5 { '

c Nuclear P:wer P! nt

References for Section 7 (Non-LOCA Transient Analysis) 1.

" Statistical Conbination of Uncertainties Methodology; Part 1:

C-E Calculated Local Power Density and Thernal tiargin/ Low Pressure LSSS for Calvert Cliffs Units I and II," CEN-124(B)-P, December,1979.

" Statistical Combination of Uncertainties Methodology: Part 2:

L i

Combination of System Paraneter Uncertain'ies in Thernal Mar 0 n Analyses for Calvert Cliffs Units I and II," CEN-124(B)-P January,1980.

" Statistical Combination of Uncertainties Methodology; Part 3:

C-E Calculated Local Power Density and Departure from Nucleate Boiling Limiting Conditiosn for Operation for Calvert Cliffs Units I and II," CEN-124(B)-P, March,1980.

A. E. Lundvall to R. A. Clark, Calvert Cliffs Nuclear Power Plant - I Docket No. 50-317. "Anendment to Operating License DPR-53 Supplement 1 2.

5th Cycle License Application," November 4,1980.

_