ML20094H235

Provides New Small Break LOCA Analysis,Per Request in Insp Repts Dtd 901026.Results Will Be Incoporated Into Next Rev to Updated SAR (June, 1992)

ML20094H235
Person / Time
Site:	Prairie Island
Issue date:	03/03/1992
From:	Parker T NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9203090191
Download: ML20094H235 (21)

Text

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

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Northem States Power Company 414 N:co!Iet Mah Minneapolis, Minnesota 55401 1t '

Telephone (612) 330 5500 March 3. 1992 U S Nutlear Regulatory Commission Attn: Document Control Desk Vashington, DC 20555 PRAIRIE IS1ANL NUCLEAR CENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Small Break LOCA Analysia. ,

We have conpleted a new Small Break Loss of Coolant Accident (LOCA) analysis.

This analysis is being subinitted per your request in Inspection Report 90015 and 90016 (dated October ?6 1990). The results of the analysis are document in Attachment 1. The andv=is is documented in the form of a draf t Updated Safety Analysis Report section (this report section is a drsft but the analysis results are final), This attachment will be incorporated into the next revision to the Updated Safety Analysis Report (June, 1992).

This accident was reanalyzed to correct an input error in auxiliary feedwater flow and several analysis concerns. The new peak cladding temperature is 107' F.

Please contact us it you have questio..s concerning these reports.

kllb%&

Themas M Parker Manager Nucleat Support Services c: Regional Administrator Region III, NRC Senior Resident Inspector, NRC NRR Froject Manager, NRC J E S11 berg -

Attachments:

1. Draft Updated Safety Analysis Report Section 14.7 0 G O D ? .>

92O3090191 920303 PDR ADOCK 0D000282 q(

30 t p PDR )f

l Prairie Island Units 1 and 2 Updated Safety Analysis Report 1

1 14.7 Lms of ReacInr Coolant.ftom SmdLEnplumU'ipes or Fr'Enfracks in Lnrag_Einn which Amrate the Emenency Core Coollr&Sntttu 14.7-1 Acceptance Cntena A minor pipe break (small break), as considered in this section, is defined as m rupture of the reattor coolant pressure boundag with a total cross-sectional area less than 1.0 ft in which the nonnally operating charging system flow is not sufficient to sustain pressurizet level rJ pressure.

This is considered a Condition III event, an infrequent fault. ,

The Acceptance Criteria for the loss-of-coolant accident is described in 10 CFR 50.46 as follows:

(a) The calculated maximum fuel element cladding temperature shall not exceed 2200 F.

(b) The calculated total oxidation'of the cladding shall nowhere exceed 0.17 times the total chdding thickness before oxidation.

(c) The calculated t Ja! an.ount of hydrogen generated from the chemical reaction of the

'hdding with water or steam shall not exceed 0.01 times the hypothetical amount that sould be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(d) Calculated changes in core geometry shall be such that the core remains amenable to cooling.

(e) After any calculated sucessful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long hved radioactivity remaining in the core.

These criteda were established to provide ignificant margin in ECCS performance following a LOCA.

14.7.2 pfelqdption of Sr"gil Break LOCA Transient

~

Ruptures of small cross-section win cause loss of the coolant at a rate which can be accomodated by the charging pumps. These pumps would maintain an operational water level in the pressurizer permitting the operator to execute an orderly shutdown. The coolant which would be released to the ,

. containment contains the fission products existing at equillboui.

The maximum break size for which the rormal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the Reactor Coolant System through the postulated break against the charging pump makeup flow at normal Reactor Coolant System pressure; i.e.,

2250 psia. A makeup flow mte fmm one charging pump is typically adequate to sustain pressurizer level at 2~250 psia for a break through a 0.375 inch diameter hole. This break results in the loss of approximately 17.25 lbm/sec.

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

Prairie Island Units 1 and 2 I Updated Safety Analysis Report i

Should a larger break occur, depressurization of the Reactor Coolant System causes Guid into the loops from the pressurizer resulting ir. a pressure and level decrease in the pressurizer. Reactor trip occurs when the low pressurizer pressure trip setpoint is reached. During the early part of the small break transient, the effect of the b~ak flow is not strong enough to overcome the Gow maintained by the reactor coolant pumps through the core as they are coasting down following reactor trip.

Therefore, upward now through the core is maintained. The Safety injection System is actuated when the appropriate setpoint is reached. The consequences of the accident are limited in two ways:

1. Control rod insertion and void foanation in'the core cause a rapid reduction of the nuclear power to a residual level corresponding to the delayed tission and fission product decay.

2. Iniection of borated water ensures sufficient flooding of the core to prevent excessive clad tempemtures.

Before the break occurt e plant is in an equilibrium condition; i.e., the heat generated in the core is being removed via the acondary system. During blowdown, heat from fission product decay, hat internals, and the vessel continues to be transferred to the Reactor Coolant System. The heat transfer between the Reactor Coolant System and the secondary system may be in either direction depending on the relative temperate.es. In the case of continued heat addition to the secondary, system pressure increases and steam dump may occur. Makeup to the secondary side is automatically provided by the auxiliary feedwater pumps. The safety injection signal stops normal feedwater now by closing the main feedwater isolation valves and initiates auxiliary feedwater flow by starting auxiliary feedwater pumps. The secondary flow aids in the reduction of Reactor Coolant System pressures.

%%n the RCS deptessurizes to the accumulator cover gas pressure, the cold leg acc mulators begin to inject water into the reactor coolant loops. Due to the loss of offsite power usumption, the ractor coolant pumps are asssumed to be tripped at the time of reactor trip during the accident and the effects of pump coastdown am included in the blowdown analyses.

14.7.3 Small Break LOCA Evajuation Model The NOTRUMP computer code is used in the analysis of loss-of-coolant accidents due to s;nall breaks in the reactor coolant system. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code ;onsisting of a number of advanced features. Among these features are the calculation of thermal non-equilibrium in all Guid volumes, flow regime-dependent

~

- drift Gux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, and regime-dependent heat transfer correlations. . The NOTRUMP small break, LOCA emergency core cooling system (ECCS) evaluation model was developed to n determine the RCS response to design basis scall break LOCAs and to address the NRC concerns expressed in NUREG-0511, " Generic Evaluation of Feedwater Transients and Small Break Loss-of Coolant Accidents in Westinghouse-Designed Operating Plants."

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_ - _ . - . - - - - - . - _ . . ,. -- - ~. . _-

Prairio Island Units 1 and 2 Updated Safety Analysis Report i

In NOTRUMP, the RCS is nodalized into volumes interconnected by flowpaths. The broken loop is modeled explicitly with the intact loops lumped in;o a second loop. The transieni behavior of the  ;

system is determined from the governing consenation equations of mass, energy, and momentum.

applied throughout the system. A detailed description of NOTRUMP is given in WCAP-10054-P-A and WCAP 10079-P-A.

The use of NOTRUMP in the analysis involves, among other things, the representation of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation. The multinode capability of the prognm enables an explicitly and detailed spatial representation of various system components. In particular, it enables 1 proper calculation 'of the behavior of the loop seal during a loss-of-coolant transient.

Citdding themial analyses are performed with the LOCTA-IV WCAP-8301-P code which uses the RCS pressure, fuel rod power history, stea:n flow past the uncovered part of the core, and mixture height history from the NOTRUMP hydraulic calculations, as input.

A schematic representation of the computer code interfres is given in Figure 14.71.

This model was developed to resolve TMI Action Item II.K.3.30. NRC acceptance of this model '

for Prairie Island was documented in an NRC staff letter dated June 6,1985.

14.7 Small Break Inout Parameters and Initial Conditions Table 14.7-1 lists important input parameters and initial conditions used in the small break analysis.

The axial power distribution and core decay power assumed for the small break analysis are shown in Figures 14.7-2 and 14.7-3.

Safety injection flow rate to the Reactor Coolant System is a function of the system pressure is used as part of the input. ' The Safety Injection System (SI) was assumed to be delivering to the RCS l 25 seconds after the generation of a safety injection signal.

For this analysis, the SI delivery considers injection fic~ which is depicted in Figure 14.7-4 as a function of RCS pressure. This figure represents injection flow from one degmded SI pump spilling

~

to either RCS pressure (if break size is smaller than SI injection line diameter), or to O psig containn"nt pressure (if break size is greater than or eaual to tise SI injection line diameter). The 25 second delay includes time requimd for diesel startup and loading of the safety injection pumps

-onto the emergency buses. Fbw from the RHR pumps does not affect the analysis since their shutoff head is lower that RB pressure during the time portion of the (nuisient co 1sidered here.

Also, minimum safeguards E aergency Core Cooling System capability and opembility has been asssumed in this analysis.

The hydmulic analyses are performed with the NCTRUMP code using 102% of the licensed core

. power. The core thermal transient analyses are performed with the LOCTA-IV code using 102% of the licensed core power.

. . ..- _ . , . - . ~.. . . _.---- . . . - . . . - . . .. - . - - ~ - .. -- .

• I Prairie Island Units 1 and 2 Updated Safety Analysis Report 14.7 5 Small Break Resulu As noted previously, the calculated peak cladding temperature resulting from a small break LOCA is less than that calculated for a large break LOCA. A range of small break analyses are presented which establishes the limiting break size. The results of these analyses are summarized in Tables 14.7-2 and 14.7-3. Figures 14.7-5 throuc;h 14.7-11 present the principal parameters of interest for the small break ECCS analyses. For the cases analyzed, the following transient parame:ers are included except hot spot clad temperature when the core is not uncovered. .

a. RCS Pressure

b. Core mixture height ,

e. Hot spot clad tempemture i For the limitng break analyzed (6 inch), the following additional transient pararneters are present (Figures 14.7-12 through 14.7-14):

- a. Core steam flow rate

b. Core heat transfer coefficient

c. Hot spot fluid temperature-

' The maximum peak clad temperature for the breaks analyzed is 1077'F. These results are well below all Acceptance Criteria limits of 10 CFR 50.46, and in no case is limiting when compared to the results presented for larne breaks.

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Prairie Island Units 'i and 2 l Updated Safety Analysis Report ]

n ,

l TABLE 14.7-1 t

INPUT PARAMETERS USED IN Tim SMALL BREAK LOCA ANALYSIS Parametel - IDPJll Core Power 107, of 1650 MWt Peak Linear Power (kW/ft) 15.096 kW/ft (includes 102% factor)

- Total Peaking Factor 2.50 Power Shape ~ See Figure 14.7-2 Fuel Assembly Array 14X14 OFA Accumula' tor Conditions:

Cover Gas Pressure 710 psig Water Volume 1250 ft3 Total Volume 2000 ft 3 Pumped Safety Injection Flow See Figure 14.7-4 Steam Generator Initial Pressure 664 psia Stern Generator Tube Plugging Level 10 %

Reactor Trip Signal 1700 psia .

- Safety Injaction Signal 1700 psia Rod Drop Time - 2.4 seccada

~

^ 2.0 seconds Reactor Trip Signal Delay Time l

l-I.

r 1

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

Prairie Island Units 1 and ?

Updated Safety Analysis Report TABLE 14.7-2 a

SMALL BREAK LOCA TIME SEQUE'4CE OF EVENTS Break Size EVENT 4.0Inen. 6.0 Inch 8.0 Inch Break Initiation, sec. 0.0 0.0 0.0 Reactor Trip Signal, sec. 8.0 5.6 5.2 Safety Injection Signal, sec. 8.0 5.6 5.2 I Top of Core Uncovered, sec. ~'60 - 140 - 78 Accumulator Injection Begins, sec. ~ 350 ~ 130 ~ 80 Peak Clad Temperature Occurs, sec.

~178 ~192 ~ 116 Top of Core Recovered, sec. ~180 -195 - 122 TABLE 14.7-3 SMALL BREAK LOCA ANALYSIS RESULTS Bmak Size RESULT ,

4.0 Inch l 6.0 Inch - l 8.0 Inch Peak Clad Temperature, *F 834 1077 1053 Peak Clad Temperature Location, ft. 10.5 10.75 10.5

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Local ZrH O 2 Reaction (max), % 0.0333 0.0339 0.0337 Local Zr/H 2O Reaction Location, ft. 10.5 10.75 10.5 Hot Rod Burst Time, sxonds NA NA NA Hot Rod Burst Location, ft. NA NA NA

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PRAIKIE ISLA O 1

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R C U i M O< TIME < CORE COVERED A P

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f Figure 14.7-1 Code interface Descriptiori for Small Break Model 4

PRAIRIE ISLAND SBLOCA . ANALYSIS Small Break LOCA Power Shape 3

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TIME (SEC) i

.:.0* SMAU. DR SAK LOC A PRAIRIEISLAND UNITS 1 AtlD 2 . - _ . .

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y-4.0* SMALL BREAK LOCA 4 CORE MIXTURE LEVEL PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.7 6 I

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TIME ISEC1 t

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6.0" SMALL BREAK LOCA PRAIRIE ISLAND UNITS 1 AND 2 I FIGl.'RE 14.7-7 l i.

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6.0* SMALL BREAK LOCA cons m aamsL i PRAIRIE ISLAND UNITS 1 AND 2 t

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TIME (S) 8.0* SMALL BREAK LOCA PRAIRlE ISLAND UNITS 1 AND 2 FIGURE 14.7 9 1

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1900. _L -

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D. 20. 40. 60. - - _ . . -80. 100. 120. 140, 160.

TIME (SEC) 8.0* CMALL BREAK LOCA PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.710 l

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0.0* SMALL CREAK LOCA i PRAIRIE ISLAND UNITS 1 AND 2 FIGURE 14.7-11

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r!.ME ISEC1 6.O* SMALL BREAK LOCA L CORE ON VAPOR FLOW PRAIRIE ISLAND UNITS 1 AND 2

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120. 140. 160. 180.

TIME (S) 6.0* SMALL BREAK LOCA HOT SPOT RCD SURFACE HEAT TRANSFER COEFFICtENT PRAIRIE ISLAND UNITS 1 AND 2 ,

FIGURE 14.7-13 i

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'O 220 2o 260 120 140 iso 180 200 TIME (S) 6.0* SMALL BR*J.AK LOCA HOT SPOT FLut0 PRAIRIE ISLAND UNITS 1 AND 2 TEMPERATURE FIGURE 14,7-14

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