ML20066C377

Nonproprietary LOFTTR2 Analysis for Steam Generator Tube Rupture for Beaver Valley Power Station Unit 2

ML20066C377
Person / Time
Site:	Beaver Valley
Issue date:	10/31/1990
From:	Holderbaum D, Robert Lewis, Schrader K WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19310D128	List: ML20066C373 ML20066C374 ML20066C377
References
WCAP-12738, NUDOCS 9101090500
Download: ML20066C377 (85)

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1 j HESTINGHOUSE CLASS 3 i

HCAP-12738 i

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LOFTTR2 ANALYSIS FOR A STEAM GENERATOR TUBE RUPTURE

) FOR BEAVER VALLEY POWER STATION UNIT 12 i

1 K. J. Schrader

D. F. Holderbaum R. N. Lewis s

C. A. Marmo l K. Rubin OCTOBER 1990 '

i l

Nuclear Safety Department I

i Hestinghouse_ Electric Corporation r Nuclear Energy Systems P.O. Box 355-Pittsburgh, Pennsylvania 15230- ,

c1990 by Westinghouse Electric Corporation l

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0644D:1D/102090 j

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TABLE OF CONTENTS i

M L

!. INTRODUCTION 1

!!. ANALYSIS OF MARGIN TO STEAM GENERATOR OVERFILL 4

}

i A. Design Basis Accident 4

B. Conservative Assumptions 5 C, Operator Action Times 7 j

0. Transient Description 13

! III. ANALYSIS OF OFFSITE RADIOLOGICAL CONSEQUENCES 25 i

} A. Thermal and Hydraulic Analysis 25

1. Design Basis Accident 25

2. Conservative Assumptions 26

3. Operator Action Times 28

! 4. Transient Description 29

5. Mass Releases 44 j B. Offsite Radiation Dose Analysis 54 IV. CONCLUSION 77 i

p V. REFERENCES .78 4

i i

(

06440:10/1 190 1

i LIST OF TABLES Idit lilla P_Ls.t 1

II.1 Operator Action Times for Design Basis Analysis 12 1

l II.2 Sequence of Events - Margin to Overfill Analysis 18 j III.1 Sequence of Events - Offsite Radiation Dose Analysis- 34 III.2 Mass Releases - Offsite Radiation Dose Analysis .50 l

III.3 Summarized Mass. Releases - Offsite Radiation Dose 51 Analysis III.4 Parameters Used in Evaluating Radiological 61 Consequences

! III.5 Iodine Spec 19c Activities in the Primary and Secondary 64 I' Coolant III.6 Iodine Spike Appearance Rates 65 4

III.7 Noble Gas Specific Activitie; in the Reactor Coolant 66 Based on 0.26% Fuel Defects i

l III.8 Atmospheric Dispersion Factors and Breathing Rates 67 III.9 Thyroid Dose Conversion Factors- 68-l' t

! III.10 Average Gamma and Beta Energy for Noble Gases 69

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06440:10/102090 11

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l LIST OF TABLES (Cont)

IAhlt Iltle Put III.11 Environmental Releases For Pre-Accident Iodine Spike Case 70 III.12 Environmental Releases for Accident Initiated 71 Iodine Spike Case III,13 Offsite Radiation Doses 72 i

0644D:10/102490 iii

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LIST OF FIGURES-E1EWit 11111 EA.21 9

!!.1 Pressurizer Level - Margin to Overfill Analysis 19 l II.2 RCS Pressure - Margin to Overfill Analysis 20 11.3 Secondary Pressure - Margin to Overfill Analysis 21 11.4 Intact Loop Hot and Cold Leg RCS Temperatures - 22-Margin to Overfill Analysis II.5 Primary to Secondary Break flow Rate - Margin to 23 <

i Overfill Analysis 6

II.6 Ruptured SG Water Volume - Margin to Overfill Analysis 24 III.1 RCS Pressure - Offsite Radiation Dose Analysis 35 i III.2 Secondary Pressure - Offsite Radiation Dose Analysis 36 j

III.3 Pressurizer Level - Offsite Radiation Dose Analysis 37 i 111.4 Rupt'u red Loop Hot and Cold Leg RCS Temperatures - 38 Offsite Radiation Dose Analysis III.5 Intact Loop Hot and Cold Leg RCS Temperatures - 39 i Offsite Radiation Dose Analysis III.6 Differential Pressure Between RCS and Ruptured 40 SG - Offsite Radiation Oose Analysis III.7 Primary to Secondary Break Flow Rate - Offsite 41 Radiation Dose Analysis 0644D:10/071890 iv  !

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3 LIST OF FIGURES '. Cont) l l

1 Figure Iltig Eagg III.8 Ruptured SG Water Volume - Offsite Radiation Dose 42 Analysis III.9 Ruptured SG Hater Mass - Offsite Radiation Dose Analysis 43 III.10 Ruptured SG Mass Release Rate to the Atmosphere - 52

j. Offsite Radiation Dose Analysis III.11 Intact SGs Mass Release Rate to the Atmosphere - 53 Offsite Radiation Dose Analysis

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III.12 Iodine Transport Model - Offsite Radiation Dose Analysis 73 III.13 Break Flow Flashing Fraction - Offsite Radiation 74 Dose Analysis III.14 SG Hater Level Above Top of Tubes - Offsite 75 i Radiation Dose Analysis 111.15 Iodine Scrubbing Efficiency - Offsite Radiation Dose 76 Analysis 1

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I. INTRODUCTION An evaluation for a design basis steam generator tube rupture (SGTR) event has been performed for the Beaver Valley Power Station Unit 2 (BVPS Unit 2) to demonstrate that the potential consequences are acceptable. This evaluation includes an analysis to demonstrate margin to steam generator overfill and an analysis to demonstrate that the calculated offsite radiation doses are less than the allowable guidelines.

The BVPS Unit 2 employs a Westinghouse pressurized water reactor (PWR) unit i rated at 2660 MWt. The reactor coolant system has three reactor coolant loops with Model 51 steam generators.

The SGTR evaluation is based on the current BVPS Unit 2 plant design which reflects the approved changes which have been incorporated since the plant was initially licensed. The evaluation is applicable for BVPS Unit 2 operation with either Westinghouse Standard or Vantage-5 Hybrid (V5H) fuel ins h11ed. The evaluation is also applicable for up to 10 percent steam generator tube plugging to provide an allowance for future tube plugging up to this level.

The SGTR analyses were performed for BVPS Unit 2 using the analysis methodology developed in HCAP-10698 (Reference 1) and Supplement 1 to WCAP-10698 (Reference 2).

The methodology was developed by the SGTR Subgroup of the Westinghouse Owners Group (WOG) and was approved by the NRC in Safety Evaluation Reports (SERs) dated December 17, 1985 and March 30, 1987. The LOFTTR2 program, an updated version of the LOFTTR) program, was used to perform the SGTR analysis for BVPS Unit 2. The LOFTTR1 program was developed as part of the revised SGTR analysis methodology and was used for the SGTR evaluation in References 1 and 2. However, the LOFTTRI program was subsequently modified to accommodate steam generator overfill and the revised program, designated as LOFTTR2, was used for the evaluation of the consecuences of overfill in WCAP-11002 (Reference 3). The LOFTTR2 program is identical to the LOFTTR1 program, with the exception that the LOFTTR2 program has the additional capability to represent the transition from two regions (steam and water) on the secondary side to a single vater regict, if overfill occurs, and the transition back to two regions again depending upon the 0644D:10/102590 1

calculated secondary conditions. Since the LOFTTR2 program has been validated against the LOFTTR1 program, the LOFTTR2 program is also appropriate for performing licensing basis SGTR analyses.-

P

, Plant response to the event was modeled using the LOFTTR2 computer code with j conservative assumptions of break size and location, condenser availability and initial secondary water mass in the ruptured steam generator. The analysis methodology includes the simulation of the operator actions.for recovery from a steam generator tube rupture based on the.BVPS Unit 2 i Emergency Operating Procedures- (EOPs), which were developed from the-Westinghouse Owners Group Emergency Response Guidelines (ERGS). In, subsequent' references to the BVPS' Unit 2 E0Ps throughout the text, the specific BVPS _

Unit 2 E0P will be listed along with the corresponding Westinghouse Owners  !

Group ERG in parenthesis.

An SGTR results in the leakage of contaminated reactor coolant-into the-secondary system and subsequent release of a portion of the. activity to the atmosphere. Therefore, an analysis must be performed to assure.that the i offsite radiation doses resulting from an SGTR are within the allowable guidelines. One of the major concerns for an-SGTR is the possibility of steam _

i generator overfill since this could potentially result in a significant l increase in the offsite radiation doses.- Therefore, an analysis was performed to demonstrate margin to steam generator overfill, assuming the limiting-single failure relative to overfill. An analysi.s was also performed to determine the offsite radiation doses, assuming the limiting single failure relative to offsite doses without steam generator overfill.- The~1imiting single failure assumptions _for those analyses are consistent with the methodology in References 1 and 2.

l l

For the margin to overfill analysis, the single failure was assupe,d to be the failure of the IheLOFTTR2 analysis to determine the margin to overfill was performed for-the time period from the steam generator tube rupture until the primary and_ secondary pressures are equalized (break flow termination). The water volume in the-secondary side of the ruptured steam generator was calculated as a function of 0644D:10/071990 2-

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time to demonstrate that overfill does not occur. The results of this i

analysis demonstrat es that there is margin to steam generator overfill for BVPS Unit 2, Since steam generator overfill does not occur, the results of the offsite radiation dose analysis represent the limiting consequences for BVPS Unit 2.

For the analysis of the offsite radiacion doses, the ruptured steam generator atmospheric steam dump valve was assumed to fail open at the time the isolation of the ruptured steam generator is performed. The primary to secondary break flow and the steam releases to the atmosphere from both the ruptured and intact steam generators were calculated for use in determining the activity released to the atmosphere. The mass releases were calculated with the LOFTTR2 program from the initiation of the event until termination of the break flow. For the time period following break flow termination, steam releases from and feedwater flows to the ruptured and intact steam generators were determined from a mass and energy balance using the calculated RCS and steam generator conditions at the time of leakage termination. The mass release information was used to calculate the radiation doses at the exclusion area boundary and low population zone assuming that the primary coolant activity is at the Standard Technical Specification limit prior to the accident. The results of this evaluation show that the offsite doses for BVPS Unit 2 are within the allowable guidelines specified in the Standard Review Plan, NURIG-0800, Section 15.6.3, and 10CFR100.

l 0644D:10/102090 3-

l l

II. ANALYSIS OF MARGIN TO STEAM GENERATOR OVERFILL An analysis was performed to determine the margin to steam generator overfill for a design basis SGTR event for BVPS Unit 2. The analysis was performed using the LOFTTR2 program and the methodology developed in Reference 1, and using the plant specific parameters for BVPS Unit 2. This section includes a discussion of the methods and assumptions used to analyze the SGTR event, as well as the sequence of events for the recovery and the calculated results.

A. Desian Basis Accident The accident modeled is a double-ended break of one steam generator tube locatedatthetopofthetubesheet _ _

he location of the break _

_ae itwas also assumed that loss of offsite power occurs at the time of reactor trip, and the highest worth control assembly was assumed to be stuck in its fully withdrawn position at reactor trip.

For the three-loop reference plant in Reference 1, the most limiting single failure with respect to steam generator overfill was determined to be a

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causing an increase in total primary to secondary leakage.

Consequently 7morewaterwillaccumulateintherupturedsteamgenerator.

For BVPS Unit 2, the RCS cooldown can be performed by releasing steam from the intact steam generators using the associated atmospheric steam dump valves or using the residual heat release valve. However, because of the

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0644D:10/102490 4

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b _ a , c, the limiting single failure for the BVPS

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Unit 2 margin to overfill anaI'ysis is assumed to be the

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, B. Conservative Assumotions Sensitivity studies were performed previously to identify the initial plant conditions and analysis assumptions which are conservative relative to stecm generator overfill, and the results of these studies were reported in Reference 1. The conservative conditions and assumptions which were used in Reference 1 were applied with the BVPS Unit 2 parameters in the LOF1TR2 analysis to determine the margin to steam generator overfill for BVPS Unit 2 with the exception of the following differences.

1. Reactor Trio and Turbine Runback A turbine runback can be initiated automatically or the operator can manually reduce the turbine load following an SGTR to attempt to prevent a reactor trip. For the reference plant analysis in Reference 1,a,c. re actor trip was calculated to occur at approximately_ a,e and turbine runback to was

_,iheeffectof simulated bas V on a runback rate of _ _ _

turbine runback was conservatively simulated by

- s,L Howeve_r, if reactor trip _ g occurs prior to turbine runback to would not be possible. It is noted that earlier reactor trip will result in earlier _ initiation of primary to secondary break flow accumulation in the ruptured steam generator and earlier initiation of auxiliary feedwater (AFH) flow. These effects will result in an increased secondary mass in the ruptured steam generator at the time of isolation since the isolation is assumed to occur at a fixed time-af ter the SGTR occurs rather than at a fixed time after reactor trip.

0644D:10/102490 5 l

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ltwou{d,beoverlyconservativetoincludetheturbinerunbackto in addition to the penalty in secondary mass due to earlier reactor trip. Thus, for this analysis, the time of reactor trip was determined by modeling the BVPS Unit 2 reactor-protection system, and turbine runback was simulated

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2. Steam Generator secondary Mass _

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^L initial see ndary water mass in the ruptured steam generator was determined by Reference 1 to oe conservative for overfill. As noted above, turbine runback was assumed to be initiated and was simulated by

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Ne initial steam generator total fluid mass was conservatively assumed to be ,

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3. AFW System Ooeration ForthereferenceplantanalysisinReference1, reactor {r[poccurred on afterthe _

SGTR, and SI was initiated on low pressurizer pressure at

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after reactor trip. The reactor and turbine trip and the assumed concurrent loss of offsite power will result in the termination of main feedwater flow and actuation of the motor + iven and turbine-driven AFW pumps. The SI signal will also result in automatic isolation of the main feedwater system and actuation of the motor-driven AFW pumps for the reference plant. For the reference plant analysis, it was conservatively assumed that AFW flow from both the turbineandmotor-drivenpugpsisinitiated _

The total AFW flow from all of the AFW pumps was assumed to be distributed uniformly to each of the steam generaters until operator actions are simulated to isolate or throttie AFW flow to control steam generator water level in accordance with the emergency procedures.

0644D:10/071890 6

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i For the BVPS Unit 2 a,nalysis, reactor trip also occurs on

,, and SI is gtiated on low pressurizer pressure at approximately after reactor trio. The time of-reactor trip and SI initiatioI was determined by modeling the BVPS Unit 2 reactor protection system, and the actuation of the AFH system

_ ~

was based on the time of Nhe flow from the turbine-driven AFH pump will be available within approximately 10 seconds-following the actuation signal, but the fic;w from the motor-driven AFH pumps will not be available until approximately 60 i seconos due to the startup and load sequencing for the emergency diesel generhtors. It was assumed that flow from both of the motor-driven AFH pumps and the turbine driven AFH pump is initiated at

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~YheAFH flow assund for the analysis is 310 gpm per steam generato7, since cavitating venturi flow elements are provided in the AFH supply lines to each steam generator which are designed to limit the flow to 310 gpm.

C. Qperator Action Times In the event of an SGTR, the operator is required to take actions to stabilize the plant and terminate the primary to secondary leakage. The operator actions for SGTR recovery are provided in the BVPS Unit 2 Plant Operating Hanual, Chapter 53A, E0P E-3 (ERG E-3), and these actions were explicitly modeled in this analysis. The operator actions modeled include identification and isolation of the ruptured steam generator,'cooldown and depressurization of the RCS to restore inventory, and termination of SI to stop primary to secondary leakage. These operator _ actions are described below.

1. Identify the ruptured steam generator.

High secondary side activity, as indicated by the air ejector discharge radiation monitor, steam generator blowdown sample radiation monitor, or main steamline radiation monitor, typically will provide the first indication of an SGTR event. The ruptured steam generator 0644D:10/102590 7

can be identified by an unexpected increase-in steam generator narrow range level, high radiation from a steam generator water sample, high radiation indication on a main steamline radiation monitor, or high radiatio" from a steam generator blowdown line. For an SGTR that q results in a high power reactor trip as assumed in this analysis, the )

steam generator water level will decrease to near the bottom of the l narrow range scale for all of the steam generators. The AFW flow will begin to refill the steam _ generators, distributing flow to each of the steam generators. Since primary to secondary leakage adds additional-liquid inventory to the ruptured steam generator, the water level will increase more rapidly in that steam generator. This response, as displayed by the steam generator water level _-instrumentation, provides confirmation of an SGTR event and also identifies the ruptured steam generator.

2. Isolate the ruptured steam generator from the intact steam generators and isolate feedwater to the ruptured steam generator.

Once a tube rupture has been identified, recovery actions begin by isolating steam flow from and stopping feedwater flow to the ruptured steam generator. In addition tn minimizing radiological releases, this also reduces the possibility of overfilling the ruptured steam generator with water by 1) minimizing the accumulation of feedwater flow and 2) enabling the operator to establish a pressure differential between the ruptured and intact steam generators as a necessary step toward terminating primary to secondary leakage. For the reference plant analysis in Reference 1, it was assumed that the ruptured steam '

generator will be isolated when i

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N r BVPS Unit 2, the steam generator narrow range level corresponding to being just on span is 5% and the comparable operator action time is 11.75 minutes. Thus, applying the Reference l' 06440:10/071890 8

methodology for the BVPS Unit 2 analysis, the ruptured steam generator was assumed to be isolated'at 27.5% narrow range level or at 11.75 minutes, whichever was longer.

3. Cool down the RCS using the intact steam generators.

After isolation of the ruptured steam generator, the RCS is cooled as rapidly as possible to less than the saturation temperature corresponding to the ruptured steam generator pressure by dumping steam from only the intact steam generators. This ensures adequate subcooling in the RCS after depressurization to the ruptured steam generator pressure in subsequent actions. If offsite power is available, the normal steam dump system to the condenser can be used to perform this cooldown. However, if offsite power is lost, the RCS is cooled using the atmospheric steam dump valves or the residual heat release valve to release steam from the intact steam generators.

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For invPS Unit 2, the atmospheric steam dump valves are

~ ~

~ ~

fi$le the residual heat release alve is

~Niieatmosphericsteamdump valves are the first alternative to perform the RCS cooldown, but if the power supply to these valves is not available, the BVPS Unit 2 E0P E-3 (ERG E-3) includes provisions to use the residual heat release valve to perform the cooldown. It is noted that a connection is provided from the steam line for each of the steam generators to the residual heat release valve, with a normally open, manual isolation valve in the connecting line. Thus, if the residual heat release valve is to be used to release steam from only the intact steam generators, an operator must be dispatched to locally close the isolation valve from the ruptured steam generator to the residual heat release valve. The BVPS Unit 2 E0P E-3 (ERG E-3) includes instructions to dispatch an operator for this purpose by the time when the ruptured steam generator -is identified and isolated. Since the time required to isolate the ruptured steam generator from the rosidual heat release valve is less than the time delay to initiate the RCS cooldown, the cooldown can be performed using either the 06440:1D/071990 9

I atmospheric steam dump valves or the residual heat release valve without any additional time penalty. Since the residual heat release valve has Idnthetwointactsteam generator atmospheric steam dump vaTves, the most limiting single failure for the margin to overfill analysis is assumed to be-the i a,c I Because offsite power is assumed to be lost and a~ sing,le failu7e of i

the as assumed for l the BVPS Unit 2 analysis, the cooldown was' performed oy dumping steam via the

- - a , e,

4. Depressurize the RCS to restore reactor coolant inventory.

When the cooldown is completed, SI flow will increa d RCS pressure until break flow matches SI flow. Consequently, SI flow must be terminated to stop primary tem.cadary leakage. However, adequate reactor coolant inventory must first be assured. This' includes both sufficient reactor coolant subcooling and pressurizer inventory to maintain a reliable pressurizer level indicatien af ter SI flow is stopped. Since leakage from the primary side will continue after SI flow ,s stopped until th: 9CS and ruptured steam generator pressures equalize, an " excess" amount of inventory is needed to ensure pressurizer level remains on span. The " excess" amount required depends on RCS pressure and reduces to zero wbon RCS pressure equals j the pressure in the ruptured steam generator.

l l

The RCS depressurization is performed using normal pressurizer spray if the reactor coolant pumps (RCPs) are running. However,-since offsite power is assumed to be lost at the time of reactor trip, the

RCPs are not running and thus normal pressurizer spray is not available. In this event, RCS depressurization can be performed using the pressurizer PORVs or auxiliary pressurizer spray. Because the pressurizer PORVs are the preferred alternative, it was assumed that a pressurizer PORV is used for the RCS depressurization for this analysis.

0644D:10/101790 10 l

5. Terminate SI to stop primary to secondary leakage.-

The previous actions will have established adequate RCS subcooling, a--  !

secondary side heat sink,-and sufficient reactor coolant inventory to ensure that SI flow is no longer needed. When these actions have'been -

l completed, SI flow must be stopped to terminate primary to secondary leakage.- Primary to secondary leakage will continue after SI flow-is stopped until RCS and ruptured steam generator pressures equalize; Charging flow, letdown, and-pressurizer heaters will then be-controlled to prevent repressurization of=the RCS and reinitiation of leakage into the ruptured steam generator. 'I Since these major recovery actions are modeled in the -SGTR analysis, it .is.

necessary to establish the times required to perform these actions.

Although the intermediate steps between the major actions are not 4 explicitly modeled,-it is also necessary to account for the time required--

to perform the steps. It hected that the total time-required. to -

comolete the recovery operations consists of both operator action time and system, or plant, response time. For instance,.the. time-for each of the major recovery operations-(i.e., RCS cooldown) is primarily due to the time required for the system response, whereas the operator action time is reflected by the time required for the operator to perform the intermediate action steps.

[ The operator action times to identify and-isolate the ruptured steam ,

generator,-to initiate RCS.cooldown, to initiate RCS depressurization, and to perform SI termination were developed-for-the design' basis' analysis in__

Reference l. Duquesne Light Company has determined the corresponding operator action times to perform these-operations for BVPS Unit 2. The-operator actions and the corresponding' operator action. times used-forLthe t

BVPS Unit 2 analysis are listed in Table II.1.-

l 0644D:10/071890 11

TABLE 11.1 BVPS UNIT 2 SGTR ANALYSIS OPERATOR ACTION TIMES FOR QESIGN BASIS ANALYSIS Action Time Intervals identify and isolate ruptured SG 11.75 min or LOFTTR2 calculated time from event initiation to reach 27.5% narrow range level in the ruptured SG, whichever is longer Operator action time to initiate 9 min

--u...- _ . _ . _ :_ ~ :2 - ::

,. _ fD#31/AMD. m_._ f*!c!d44M-D N W MR3 Operator action time to initiate 2.5 min depressurization Depressurization Calculated by LOFTTR2 Operator action time to initiate 1.25 min SI termination l

SI termination and pressure Calculated by LOFTTR2 equalization 0644D:10/071890 12

l l 0. Transient Descriotion The LOFTTR2 analysis results for the BYPS Unit. 2 margin to overfill- s analysis are described below. The sequence of events for this transientL i

is presented in Table II.2. A Following the tube rupture, reactor coolant flows ~ from the primary _into' j

the secondary side' of the ruptured steam generator _sinceithe primary i j pressure is greater than-the steam generator pressure. In response to

' j this loss of reactor coolant,; pressurizer level decreases as shown in- '

Figure 11.1. The RCS pressure also decreases as shown in Figure II.2 as the steam bubble in _the pressurizer expands. As the RCS. pressure decreases due to the continued. primary to secondary leakage, automatic '

reactor trip. occurs on'a low pressurizer pressure trip signal at

_ _ accroximately.lBa sece::dt. _

~

j After reactor trip, core power rapidly decreases to_decayleJLLlents.

~

ine turuine stop valves close and steam flow to the turbine is terminated. The steam dump system is designed.to actuate following reactor trip to limit the increase in secondary pressure,' but the steam 1

dump valves remain closed due to the loss of- condenser vacuum resulting from the assumed loss of offsite power at the time of. reactor trip. Thus, the energy transfer -from the primary -system causes' the- secondary side pressure to increase rapidly after reactor trip until the steam-generator--

atmospheric steam dump valves (and safety valves if theirLsetpoints are reached) lift to dissipate the energy, as shown in Figure 11.3.-

l l

' The pressurizer level and RCS pressure decrease more rapidly after. reactor trip as energy transfer to the secondary shrinks-the reactor coolant and

-the leak flow continues to deplete primary inventory. The. decrease in RCSi -

i inventory results in a low pressurizer pressure SI-signal at approximately L

161 seconds. The main feedwater flow will be isolated and AFW flow will .

be automatically initiated following SI actuation. After SI-actuation,

-the SI flow rate initially exceeds the tube rupture break flow rate, and the pressurizer level and RCS pressure begin to increase ~and subsequently stabilize at the equilibrium values where the SI flow rate equals the break flow rate. ,

i 06440:10/071990 13

Since offsite power is assumed lost-at reactor trip, the RCPs trip and a gradual transition to natural circulation flow occurs. Immediately following reactor trip the temperature differential across the core decreases as core power decays (see Figure ll.4);1however, the temperature differential subsequently increases as the reactor-coolant. pumps coast -

down and natural circulation flow-develops. The cold leg temperatures trend toward the steam generator-temperature as the fluid residence _ time in the tube region increases. The hot leg temperature reaches a peak and-then slowly decreases, as steady-state conditions are reached, until operator actions are initiated to cool'down the RCS.

Major Operator Actions

1. Identify and Isolate the Ruptured Steam Generator 4 _

Once a tube-rupture-has been identified, recovery actions begin by.

isolatino steam flow from the rup.lgadatumJMRtit9.rJad_150htina the auxiliary feedwater flow to.the ruptured steam generator. As indicated previously, the ruptured steam generator is assumed to be identified and isolated when the narrow range -level reaches.27.5% on the ruptured steam generator er at 11;75 .tr.utc; Ofter initi:ti:n of the SGTR, whichever is longer. For the BVPS Unit 2 analysis, the time-to reach 27.5% is less than 11.75 minutes, and thus the ruptured steam generator is assumed to be isolated at 11.75 minutes. However, because of the computer program limitations for simulating the operator actions, the ruptured steam generator was isolated three seconds-later at 708 seconds.

2. Cool down the RCS to Establish Subcooling Margin-After isolation of the ruptured steam generator, -there .is a 9-minute operator action time imposed-prior to initiating the cooldown.-lThe actual delay time used in the analysis. is 2 seconds longer because of the computer program limitations-for simulating the operator actions.

After this time, actions are taken to cool th'e RCS as rapidly as.

possible by dumping steam from the intact steam generators. Since.

offsite power is lost, and the limiting single failure is the failure 0644D:10/071890 14 9

~ '

.of the the RCS is

~ '

cooled by dumping steam to the atmosphere using the

..y YThe"

_a -

are assumed to be opened at-1250 seconds for RCS cooldown. The coidown is continued until RCS subcooling at the ruptured steam generator pressure is 20*F plus an allowance for subcooling uncertainty. When these conditions are satisfied at 2098 seconds, it is assumed that the operator closes the-e., c.

terminate the cooldown. This cooldown ensures that there will be adequate subcooling in the RCS after the subsequent depressurization of the RCS to the ruptured steam generator pressure. The reduction in the intact steam generator pressures required to accomplish the cooldown is shown in Figure 11.3, and the effect of the cooldown on the RCS temperatuie is shown in Figure 11.4'. The pressurizer level '

decreases during this cooldown preciiss due to shrinkage of the reactor - - -

coolant a63 goes offscaTe TW Ti'ihoin in Figure 11.1. The RCS pressure also decreases due to the RCS cooldown as shown in Figure 11.2.

) 3. Depressurize RCS to Restore Inventory After the RCS cooldown, a 2.5 minute operator action time is included prior to the RCS depressurization. The actual-delay time used in the dnalysis is tWO seconds longer because of the computer program limitations for simulating operator actions. The RCS depressurization is performed to assure adequate coolant inventory prior.to terminating SI flow. With the RCPs. stopped, normal pressurizer spray is not available and thus the RCS is depressurized by opening a pressurizer PORV. The RCS depressurization is initiated at 2250 seconds.and continued until any of the following conditions are satisfied:

pressurizer level is greater than 76%, or RCS subcooling is less than the allowance for subcooling uncertainty, or RCS pressure is less than the ruptured steam generator pressure and pressurizer level is greater than 4%. For this case, the RCS depressurization is terminated because the RCS pressure is reduced to less than the ruptured steam 0644D:10/071990 15

generator pressure and the pressurizer level is above 4%. The RCS depressurization reduces the break flow as shown in Figure II.5, and increases SI flow to refill the pressurizer as shown in Figure 11.1.

4. Terminate SI to Stop Primary to Secondary Leakage The previous actions establish adequate RCS subcooling, a secondary side heat sink, and sufficient reactor coolant inventory to ensure that SI flow is no longer needed. When these actions have been completed, the SI flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary leakage. The SI flow is terminated at this time if RCS subcooling is greater than the allowance for subcooling uncertainty, minimum AFW flow is available or at least one intact steam generator level is in the narrow range, the RCS pressure is stable or increasing, and the pressurizer level is i greater than 4%. For the BVPS Unit 2 analysis, SI was not terminated

~ ~ ~

until the RCS pressure increased to 50 psi above the ruptured steam generator pressure to assure that RCS pressure is increasing.

After depressurization is completed, an operator action time of 1.25 minutes was assumed prior to initiation of SI termination. Since the above requirements are satisfied, SI termination actions were performed at this time by closing off the SI flow path. After SI i termination, the RCS pressure begins to decrease as shown in

~

Figure 11.2. The

~ ~

are also opened to dump steam to maintain the prescribed RCS temperaturetoensurethatsubcoolingismgigtained. When the tlieincreasedenergy transfer from primary to secondary also aids in the depressurization of the RCS to the ruptured steam generator pressure. The primary to secondary leakage continues after the SI flow is terminated until the RCS and ruptured steam generator pressures roralize.

l l

06440:10/071890 16 l

The primary to secondary break-flow rate throughout the recovery operations is presented in Figure.!!.5. The water volume in the ruptured steam generator is presented as a function of time in Figure II.6. It is noted that the water volume.in the ruptured steam generator when the break flow is terminated is less than the total steam generator volume of 57593 ft . Therefore, it is concluded that overfill of the ruptured steam generator will not occur for a design basis SGTR for BVPS Unit 2.

06440:10/071890 17

TABLE !!.2 BVPS UNIT 2 SGTR ANALYSIS

~

SEOUENCE OF EVENT!,

MARGIN TO OVERFILL ANALYSIS EVENT Time (sec)

SG Tube Rupture 0 Reactor Trip 154 Safety Injection 161 Ruptured SG Isolated 708 i RCS Cooldown Initiated 1250 RCS Cooldown Terminated 2098 RCS Depressurization Initiated 2250 RCS Depressurization Terminated 2364 SI Terminated 2440 i

Break Flow Terminated 2880:

l l

0644D:10/071990 18

I BEcvER VALLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE PRESSURIZER LEVEL

50. J BC.

70, b60.

W 5 E0.'

u R

Q40.

G L*)

E50.

ii 20.4 l

0.-

c. I

0. 500. 1000. 1500. 2000. 2500. 5000. 5500.

TIFC (SCC)

Figure II.1 Pressurizer Level - Margin to Overfill' Analysis-l 06440:10/071890 19

ll i BEcvER vcLLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE RCS PRESSURE

?400.-

2200.

2000.-

2

{1600,. & /

W.

1600, e

y:t::..

I203.

m 1000..

l 900.O. 500. 1000. 1500. 2000. 2500. 5000. 5500.

TIMC (SECl Figure II.2 RCS Pressure - Margin to 0verfill- Analysis 0644D:10/072890 20

EEAVER VALLEY UNIT 2 STEAM OENERATOR TUBE RUPTURE SECONDARY PRESSURE 1400.4

.1200.

RUPTURED SG 5:000.-

f_

N 000. -

s

! INTACT SG g 600.-

i

) SOO.

200.

0 *0. 500. 1000. 1500. 2000. -2500, 5000. 5500.

TIME tSEC1 l

Figure !!.3 Secondary Pressure - Margin to Overfill Analysis 06440:10/071890 21

r.

i SEAVER VALLEY UNIT 2 STEAM OENERATOR TUBE RUPTURE INTACT LOOP HOT AND COLD LEO RCS TEt1PER ATURES 650.-

600.

THOT 550.-j g TCOLD 500.

c 3

,450,-

G

=

$400.

3

!!O.

I

l. E00 0. 500. 1000. 1500. 2000. 2500. 5880. 5500.

' TIME ISEC)

I Figure II.4 Intact Loop Hot and Colo Leg RCS Temperatures -

Margin to Overfill Analysis .

1 06440:10/071890 22 l

SEAVER VALLEY UNIT 2 STEAN GENERATOR TUBE RUPTURE.

PRIMARY TO SECONDARY BREAK FLOW

• C..

90.

70.-

60. ,-

u

~4 h 50.

i E 40..

d M 50.

b 20 1

0.-

O.

10.O. 500. 1000. 1500. 2000. E500. 5000. !T00,

?!MC tSCO)

Figure II.5 Primary to Secondary Break Flow Rate -

Margin to Overfill Analysis 0644D:10/071890 23

BEAVER VALLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE RUPTURED SG WATER VOLUME 6000.-

MODEL $1 SG SECONDARY VOLUME 5500.

5 5000.-

E 5 4500,f 5

a g 4000.

=

E (3500.

E I

!000.!

l

! 25000. 500. 1000. 1500. 2000, 2500. 5000. 3500.

-TINC (SC01 Figure II.6 Ruptured SG Water. Volume - Margin to Overfill Analysis 1

0644D:10/071890 24

III. ANALYSIS OF OFFSITE RADIOLOGICAL CONSEQUENCES An analysis was performed to determine the offsite radiological consequences for a design basis SGTR for BVPS Unit 2. The thermal and hydraulic and- the offsite radiation dose analyses were performed using the methodology developed in References 1 and 2 and the plant specific parameters for BVPS Unit 2.

A. Thermal and Hydraulic Analysis The plant response, the integrated primary to secondary break flow and the mass releases from the ruptured and intact steam generators to the condenser and to the atmcsphere until break flow termination were calculated with the LOFTTR2 program for use in calculating the offsite radiation doses. This section provides a discussion of the methods and assumptions used to analyze the SGTR event and to calculate the mass releases, the sequence of events during the recovery operations, and the calculated results.

1. Desian Basis Accident The accident modeled is a double ended break of one steam generator tube located at the top of the tube sheet on the outlet (cold leg) side of the steam generator.

~

e.

-s,lt was also assumed that loss of offsite power occurs at the time of reactor trip and the highest worth control assembly was assumed to be stuck in its fully withdrawn position at reactor trip.

Based on the information in Reference 2, the most limiting single failure with respect to offsite doses for BVPS Unit 2 is'a failed open atmospheric steam dump valve on the ruptured steam generator. Failure of this atmospheric steam dump valve will cause an uncontrolled depressurization of the ruptured steam generator which will increase 0644D:10/071890 25

?

primary to secondary leakage and the mass release to the atmosphere.

Pressure in the ruptured steam generator will remain belcw that in the primary system until the failed atmospheric steam dump valve is isolated by locally closing the associated block valve, and the recovery actions are completed. Thus, for the offsite dose analysis, it was assumed that the ruptured steam generator atmospheric steam dump valve fails open and must be-locally isolated.

2. Conservative Assumotions Most of the conservative conditions and assumptions used for the margin to overfill analysis are also conservative for the offsite dose  :

analysis, and thus most of the same assumptions were used for both analyses. The major differences in the assumptions which were used for the LOFTTR2 analysis for offsite doses are discussed below,

a. Reactor Trio and Turbine Runback An earlier reactor trip is conservative for the offsite dose analysis, similar-to the case for the overfill analysis. Due to the assumed loss of offsite power, the condenser is not available for steam releases once the reactor is tripped. Consequently, after reactor trip, steam is released to the atmosphere through-the steam generator atmospheric steam dump valves (and safety valves if their setpoints are reached).' Thus, an earlier trip time leads to more steam released to the atmosphere from the ruptured and intact steam generators. The time of the reactor trip was calculated by modeling the BVPS Unit 2 reactor protection

_ system, and this time was used for the offsite dose analysis.

a,c 06440:10/071990 26 L

b. Steam Generator Secondary Mass

- 1 If steam generator overfill does not occur, a

~Iesultsinaconservative prediction of offsite doses. Thus, for the offsite dose analysis, the initial secondary mass was assumed to cogrpspond to operation at nominal steam generator mass minus a

~ ~~

allowancefor uncerteinties. As noted above, a , t.

c. AFW System Operation

~

~

In Reference 2, it was determined that a results in an increase in the calculated offsite radiation doses _

for an [GTR, whereas it was previously concluded that

~

fs conservative for the margin to overfill analys"Is.

~

However, it was also demonstrated in Reference 2 that a

-n Since the single failure assumed for the

~

offsite radiation dose analysis is a failed open atmospheric steam dump valve on the ruptured steam generator, it-is not necessary_to assume an additional failure in the AFW system.- Thus, the turbine-driven pump and both motor-driven pumps were assumed to deliver flow to the three steam generators, and an AFW flow of 310 gpm per steam generator was assumed for the offsite radiation dose analysis. The delay time assumed for delivery of the AFW flow was conservatively {

a , c.,

06440:10/071890 27 t

1

d. Flashina Fraction I

When calculating the amount of break flow that flashes to steam, 100 percent of the break flow is assumed to come [ rom the hot leg side of tha break sincethetube rupture flow actually consists of flow from the hot leg and cold leg sides of the steam generator, the temperature of the~ combined flow will be less than the hot leg temperature and the flashing fraction will be correspondingly lower. Thus the assumption that 100 percent of the break flow comes from the hot leg is conservative for the SGTR analysis.

3. Ooerator Action Times The major operator actions required for the recovery from an SGTR are discussed in Section II.C and the operator action times used for the overfill analysis are presented in Table 11.1. The operator action times assumed for the overfill analysis were also used for the offsite dose analysis. However, for the offsite doses analysis, the atmospheric steam dump _ valve on the ruptured steam generator was assumed to fail open at the time the ruptured steam generator is isolated. Before proceeding with.the recovery operations, the failed a ^n atmospheric steam dump valve on the ruptured steam. generator is assumed to be isolated by locally closing the associated block valve.

Duquesne Light Company has determined that an operator can locally close the block valve.for the atmospheric steam dump valve on the ruptured steam generator within 6.5 minutes after the failure. Thus, it was assumed that the ruptured steam generator atmospheric steam dump valve is isolated at 6.5 minutes after the valve -is assumed to fail open. - After the ruptured steam generator atmospheric steam dump valve is isolated, the additional delay time of 9 minutes (Table 11.1) was assumed for the operator action time to initiate the RCS cooldown.

l i

l l

06440:1D/071890 28 l

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

4. Transient Descriotion The LOFTTR2 analysis results for the offsite dose evaluation are described below. The sequence of events for the analysis of the offsite radiation doses is presented in Table'III.1. It is noted that reactor trip occurs at a slightly different time for this case compared to the overfill analysis due to the use of different input parameters to provide conservative results for the offsite dose analysis. The transient results for this case are similar to the transient results for the overfill analysis until the ruptured steam generator is isolated. The transient behavior is different after this time since it is assumed that the ruptured steam generator atmospheric steam dump valve fails open at that time.

Following the tube rupture the RCS pressure decreases as shown in Figure III.1 due to the primary to secondary leakage. In response to this depressurization, the reactor trips on low pressurizer pressure at approximately 159 seconds. Af ter reactor trip, core power rapidly decreases to decay heat levels and the RCS depressurization becomes more rapid. The steam dump system is inoperable due to the assumed loss of offsite power, which results in the secondary pressure rising to the steam generator atmospheric steam dump valve-setpoint as shown in Figure III.2. The RCS pressure and pressurizer level also decrease more rapidly following reactor trip as..shown in Figures III.1 and 4 III.3. The decreasing pressurizer pressure leads to an automatic SI signal on low pressurizer pressure at approximately 169 seconds.

Major Operator Actions

1. Identify and Isolate the Ruptured Steam Generator The ruptured steam generator is assumed to be identified and isolated at 11.75 minutes after the initiation of the SGTR or when tha narrow range level reaches 27.5%, whichever time is greater.

Since the time to reach 27.5% narrow range level is greater than 11.75 minutes, it was assumed that the ruptured . steam generator is 06440:10/102090 29

i isolated when the level reaches 27.5% which occurs at 808 seconds.

The ruptured steam generator atmospheric steam dump valve is also assumed to fail open at this time, and the failure is simulated at 810 seconds because of the computer program limitations. The failure causes the ruptured steam generator to rapidly depressurize, which results in an increase in primary to secondary leakage. The depressurization of the ruptured steam generator increases the break flow and energy transfer from primary to secondary which results in a decrease in the ruptured loop temperatures as shown in Figure III.4. The intact steam generator loop temperatures also decrease, as shown in Figure 111.5. It is assumed that the time required for the operator to identify that the ruptured steam generator atmospheric steam dump valve is open and to locally close the associated block valve is 6.5 minutes. However, the actual time used in the analysis is 2 t

seconds longer becausa of the computer program limitations. Thus, at 1202 seconds the depressurization of. ruptured steam generator is terminated and the ruptured steam generator pressure begins to increase as shown on Figure III.2.

2.

Cool Down the RCS to establish Subcooling Margin After the block valve for the ruptured steam generator atmospheric steam dump valve is closed, there is a 9 minute operator action time imposed prior to initiation of cooldown. Thus, the RCS cooldown was initiated at 1742 seconds. By this time, the ruptured steam generator pressure nas increased to the intact steam generator pressure and stablized at that value. :The RCS cooldown target temperature is determined based on the ruptured steam generator pressure at that time. Since offsite power is lost, the RCS is cooled by dumping steam to the atmosphere using the intact steam generator atmospheric stuam dump valves. The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20'F plus an allowance for instrument uncertainty. Because the ruptured steam generator pressure has 1

1 0644D:1D/071990 30 I

increased to the intact steam generator pressure prior to performing the cooldown, the associated temperature the RCS must be cooled to is not as. low, which has the net effect of reducing-the time required for cooldown.

The cooldown is initiated at 1742 seconds and is completed at 2292 seconds.

The reduction in the intact steam generator pressures required to accomplish the cooldown is shown in Figure III.2, and the effect of the cooldown on the RCS temperature is shown in Figure III.S.

The RCS pressure and pressurizer level. also decrease during this cooldown process due to shrinkage of the reactor coolant as shown in Figures III.1 and III.3.

3. Depressurize RCS to Restore Inventory After the RCS cooldown, a 2.5 minute operator action time is included prior to the RCS depressurization. The RCS is depressurized to assure adequate coolant inventory prior to terminating SI flow. With the RCPs stopped, normal pressurizer spray is not available and thus the RCS is depressurized by opening a pressurizer PORV. The RCS depressurization_is initiated at 2442 seconds and continued until.any of the.following conditions are satisfied: pressurizer level is greater than 76%,

or RCS subcooling is less than the allowance for subcooling uncertainty, or RCS pressure is less than tha ruptured steam generator pressure and pressurizer level is greater than 4%. For this case, the RCS depressurization is terminated because the RCS pressure is reduced to less than the ruptured steam generator pressure and the pressurizer level is above 4%. The RCS depressurization reduces the break flow as shown in Figure III.7, and increases SI flow to refill the pressurizer as shown in Figure III.3, t

0644D:10/071990 31 l

l

4. Terminate SI to Stop Primary to Secondary Leakage- f The previous actions establish adequate RCS _subcooling, a secondary side heat sink,:and sufficient _-reactor coolant inventory-to ensure that SI flow is no longer needed. When these actions ,

have been completed, the SI flow must be stopped to prevent repressurization of the RCS and-to terminate primary to-secondary 3 leakage. The SI flow is-terminated at this time'if RCS subcooling-is greater than the allowance for:subcooling uncertainty.. minimum ,

AFW flow is available or at-least_one intact steam generator level 4 is in the narrow range, the RCS pressure is stable or increasing, and the pressurizer level is greater than 4%. For the BVPS Unit 2  !

analysis, SI was not terminated until the RCS. pressure increased to 50 psi above the ruptured steam generator pressure _to assure  ;

thN RCS pressure is increasing.

After'depressurization'is_ completed, an operator. action time of 1.25 minutes was assumed prior to initiation of SI termination.-

Since the above requirements are satisfied, SI termination actions-were performed at this time -by closing off:the-SI flow path.

After SI termination, the RCS pressure begins to decrease as shown in Figure III l.- The intact steam generator atmospheric steam- ')

dump valves are also opened to dump steam to maintain the prescribed RCS temperature to' ensure that subcooling is maintained. When the atmospheric steam dump valves are opened, the increased energy transfer from_ primary to secondary also aids in the depressurization of the'RCS to'the ruptured ' steam generator pressure. The differential-pressure between'the RCS and the 4 ruptured steam generator is shown_in Figure III.6. _ Figure III.7-shows-tnat the primary to secondary leakage continues after the SI flow-is stopped until_ the RCS and ruptured steam generator pressures-equalize.

The ruptured steam generator water volume is shown in Figure III.8.

For this case, the water volume in the ruptured steam generator when a the break flow is terminated is less than the volume for the margin to d

0 1

0644D:10/071890 32 l l

I l l

l i F overfill case and significantly less than the total steam generator

]

volum of 5759 ft 3. The mass of water in the ruptured steam

! generator is also shown as a function of time in Figure !!!.9.

i i

i i

i 1

i j

i l

2 i

i i

1 j

i 4

i k

l.

1 l

l 0644D:lD/071890 33 i i

,v.,-, - ,..,.. ,. _ , ,- , - . . , ,,.c y...,- -- - . _ . - . ~ ,r,-.,, -~. . - - - , .....~,,....._.---.-._..m.-,.-,.- - - . - , .. --,4

\ .

l 1

i l TABLE 111.1 j BVPS UNIT 2 SGTR ANALYSIS

SE00ENCE OF EVENTS l OFFSITE RADIATION DOSE ANALYSIS g TIME (sec) 1 SG Tube Rupture 0 Reactor Trip 159 i

Safety injectiore 169 l Ruptured SG isolated 808 Ruptured SG Atmospheric Steam Dump Valve Feils Open 810 g Ruptured SG Atmospheric $ team Dump Valve Blo:k Valve Closed 1202 RCS Cooldown Initiated 1742 RCS Cooldown Terminated 2292-RCS Depressurization Initiated 2442 RCS Depressurization Terminated 2558 S1 Terminated 2634 l Broak Flow Terminated 3070 1

(

06440:10/071890 34 i

EEAVER VALLEY 'J N I T 2 STEAM GENERATOR TUBE RUPTURE RCS PRESSURE 2422 1 2200.-

22C2.

i

• F G

ggg, e (" %

h1622.<

I e

1422.'

1220.

g g ', '. -

't . 500. 2000. 1500. 2000. 2500. 5000. !500.

TIMC (SC: 1 Figure !!!.1 RCS Pressure - Offsite Radiation Oo=e Analysis 0644D:10/071890 35

i EEAVER vcLLEv UNIT 2 STEAM OENERATOR TUBE RUPTURE SECONDARY PRESSURE

'400.?

):22.

1 RUPTub(D M

.ppp I- 6 -' ~ ^

e /

'k W

' ' RUPTURED $G c

Etc.'

=

E '

INTACT SG

7. sec "

see, U 'g , 500, lege, 1500, 2000. 2500. 5809. 1922.

ttat ist;)

s Figure 111.2 Secondary Pressure Offsite Radiation Dose Analysis z 06440:10/071890 ,

36

i ll 1

i i

1 BEAVER VALLEY UNIT a STEAM GENERATOR TUBE RUPTURE i

PRESSURIZER LEVEL 4

1

! 60.<

i 70.

1 l .

60..

U

50, 5

a 1 a '

• se.

W

.f 4 o l

^ 50.

. .F.

i 20.t i

I w" a

10. -e

, ,. ,s v ,,.,

O .0. 500.

- ^

1568. 2008, 2589. 5MO. 5500.

1988.

flMC ISCC1 l Figure !!!.3 Pressurizer Level Offsite Radiation Dose Analysis c

) '

s y .n ,

_06440:10/071890 37 d

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

i 1

BEAVER VALLEY UNIT 2 STEAN GENERATOR TUBE RUPTURE i

RUPTURED LOOP HOT AND COLO LEO RCS TEMPERATURES 650."

600. .

• TH0T 5 gga j h,

TCOLD 500 .

E E

E d50..

3 k400.

_5 550, 500 0- 500. 1000. 1500. 2000. 2500. 5000. 5500.

T!MC (SCO)

Figure 111.4 Ruptured Loop Hot and Cold Leg RCS Temperatures -

Offsite Radiation Dose Analysis l

06440:10/071890 38

l i

EEavER VALLEY UNIT 2 $7EAM GENERATOR TUBE RUPTURE INTACT LOOP HOT AND COLD LEO RCS TEt1PERATURES 650.'

600.-

TH0T 3 -

w if a. j' g TCOLD E 500.

U

=

5 450.-

2

=

400.

E

!50.

500,O. 500, 1000. -1500. 2000. 2500. 5800. 5500, t!NC (SCCI Figure 1.11.5 Intact Loop Hot and Cold leg RCS Temperatures -

Offsite Radiation Dose Analysis t

06440:10/071890 -39

, , , . , .,.,,.y. - -

r., - . , - - _ . - - - . - _ . - - - . . - _ . ,.m- -.

-.,y ,~ , r -----er.- - - y _-,-,m.,~,, u

5 l

l l

BEAVER VALLEY UNIT 2 STEAM OENERATOR TUBE RUPTURE

!~

DIFFERENTIAL PRESSURE BETWEEN RCS AND RUPTURE? SG

! 1600.*

Idee.-t 1200.<

1000..

O

~

C p.600.'

g 600.

400..

200.

O.

'200 8. 500. 1000. 1500. 2000. 2500. 5008, 5500, f!NC (SCO)

Figure !!!.6 Differential Pressure Between RCS and Ruptured SG -

Offsite Radiation Dose Analysis 06440:10/071890' 40

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

1 i

1

)

! BEcvER VALLEY UNIT 2 STEAM OENERATOR TUEE RUPTURE

}

l,1 P R I M A Pl'70-iC0GiOMP~ 6 RCACTGM ~-'-~1~-"~~'~~~~

l t

j 8

S0. t 60.-

t j 70.'

(O. --

' v N ,

g VC."

s 4

E 40,

.a

' $ (0.4 r-1 20.'

0.'

O. }

10'O. 500. 1000. 1500. 2006. 2500. 5000. 5500, i

TIMC iSC 1 1

Figure !!!.7 Primary to Secondary Break Flow Rate -

Offsite Radiation Dose Analysis 1

3 4

06440:10/071890 41 c

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

i i

1 1

i BEAVER VALLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE

\

, . . , , RUPT[lp!D $O L'A7ER t'ai,;'m, ' '

1 1220.-

1 4

b t 4000, -

f I !!00.<

=

5 g 5000.<

=

E r= 2500.

1 G

2020. ]

sec.O. 600. late. 1500, 2000. 2500. 5000. 5500.

TIMC ISCCl-figure !!!.8 Ruptured SG Water Volume Dose Analysis Offsite Radiation 06440:10/071890 42 J

.. ,,,.,u_._,., .,,-,.cwe-ar, c , - . , , . . , , , . , , . . , ,.-.m.,,,v- y--,-_.co. -,,,,.,-..,,.,,,,,f,_ww.rw#

y_. , ,_w,,mm__, , _ . . . + -

y 7-3.#_,.,--c,..,,-.,---

i 1

l 1

J i

SEAVER VALLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE RUPTURED SG WATER NA55 1

240000.-

' 220000.

5 200000.>

d m

j160000.- . w e* q - As a :., +. , ,a 9 . ,,. .

m

=

5 160000..

i 8

e r 140000,.

h

- l20000,4 100000.'

60000.O. 500. 1000. 1508, 2000. 2598. 5000. 3508.

TIMC ISCC)

Figure !!!.9 Ruptured SG Water Mass Offsite Radiation Dese Analysis 06440:10/071890- 43 1-1-

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

i S. Mass Releases The mass releases were determined for use in evaluating the exclusion

>,m..,_. area boundary and low ooonlation tnSe evii1 tion evorure. The steam r,elease,s

-. . fynm the runtured nad in+e.-t rt .m .aen,a mreer, th. f r 6fpp-flows' to the ruptured and intact steam generators, and primary to

~

secondary break flow into the ruptured steam generator were determined for the period from accident initiation until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident and from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident. The releases for 0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are used to calculate the radiation doses at the exclusion area boundary for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exposure, and the releases for 0 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are used to calculate the radiation doses at the low population zone for the duration of the accident.

.. .. . -_. . .. . . .. . . J a._th.* J MTTP.2. ?.W,.Wtu. the MTD.mer.g ae t ions in BVPS Uni t 2 E0P E 3 (ERG E 3) were simulated until the termination of primary to secondary leakage. After the primary to secondary leakage is terminated, the operators will continue the SGTR recovery actions to prepare the plant for cooldown to cold shutdown conditions. When these recovery actions are completed, the plant should be cooled and depressurizedtocoldshutdogw(conditions, it was assumed that the cooldown is performedusingBVPSUnitIE0 PES 3.3(ERGES-3.3),POSTSGTR C00LDOWN VSING STEAM DUMP, since this method results in a conservative evaluation of the long term mass releases for the offsite dose analysis.

The high level actions for the post SGTR cooldown method using steam dump in BVPS Unit 2 E0P ES-3.3 (ERG ES-3.3) are discussed below.

1. Prepare for Cooldown to Cold Shutdown The initial steps to prepare for cooldown to cold shutdown will be continued if they have not already been completed. A few additional steps are also performed prior to initiating cooldown,

06440

10/071890 44 i

i These include isolating the cold leg Si accumulators to prevent unnecessary injection, energizing pressurizer heaters as necessary to saturate the pressurizer water and to provide for better pressure control, and assuring adequate shutdown margin in the ,

event of potential boron dilution _due to in leakage from the ruptured steam generator.

2. Cooldown RCS to Residual Heat Removal (RHR) System Temperature The RCS is cooled by steaming and feeding the intact steam  ;

generators similar to a normal cooldown. Since all immediate safety concerns have been resolved, the cooldown rate should be maintained less than the maximum allowable rate of 100'F/hr. The preferred means for cooling the RCS is steam dump to the condenser p since this mjpjmires the radiological rp1pases and conserves feedwater supply. The atmospheric steam dump valves for.the intact steam generators can also be used if steam dump to the condenser is unavailable. Since a loss of offsite power is assumed for the BVPS Unit 2 analysis, it was assumed that the cooldown is performed using steam dump to the atmosphere via the intact steam generator atmospheric steam dump valves. When the RHR system operating temperature is reached, the cooldown is stopped until RCS pressure can also be decreased. This ensures that pressure / temperature limits will not be exceeded.

3. Depressurize RCS to RHR System Pressure When the cooldo*- to'RHR system temperature is completed, the pressure in the ruptured steam generator is decreased by releasing steam from the ruptured steam generator. Steam release to the condenser is preferred since-this minimizes radiological releases, but steam can be released to the atmosphere using the atmospheric steam dump valve on the ruptured steam generator if the condenser is not available. Consistent with the assumption of a loss of-offsite power, it was assumed that the ruptured steam generator is 06440:10/071890 45

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

l 1 depressurized by reinsing steam via the atmospheric steam. dump f valve. As the ruptured steam generator pressure is reduced, the l RCS pressure is maintained equal to the pressure in the ruptured +

i steam generator in order to prevent in leakage of secondary side water or additional primary to secondary leakage. Although normal pressurizer spray is the preferred means of RCS pressure control, auxiliary spray or a pressurizer PORY can be used to control RCS pressure if pressurizer spray is not available.

4. Cooldown to Cold Shutdown i

When RCS temperature and pressure have been reduced to the RHR system in service values, RHR cp h n cooling is inittsted to i

complete the cooldown to cold shutdown. When cold shutdown conditions are achieved, the pressurizar can be cooled to terminate the event.

2 The methodology in Reference 2 was used to calculate the mass releases for the BVPS Unit 2 analysis. The methodology and the results of the calculations are discussed below.

4. Methodology for Calculation of Mass Releases 4

The operator actions for the SGTR recovery up to the termination of primary to secondary leakage are simulated in the LOFTTR2 analyses. Thus, the steam releases from the ruptured and intact steam generators, the feedwater flows to the ruptured and intact steam generators, and the primary to secondary leakage into the ruptured steam generator were determined from the LOFTTR2 results for the period from the initiation of the accident until-the 1cakage is terminated.

Following the termination of leakage, it was assumed that the RCS andintactsteamgegeratorconditionsaremaintainedstablefora until the cooldown to cold shutdown is initiated. The atmospheric steam dump valves for the intact steam 0644D:10/071890 46

i generators were then assumed to be used to cool down the RCS to j the RHR system operating temperature of 350*F, at the maximum allowable cocidown rate of 100'F/hr. The RCS and the intact steam generntor tornporstn*ns r.t f, Sn'. err. '.nirs thor tiet9mine(

~

.T.he_ steam releases and the feedwater flows for the intact steam enerator for the period from leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were determined from ,

! - a,e

~

Since the ruptured steam generator is isolated, no change in the ruptured steam generator conditions is assumed to occur until subsequent depressurization.

The RCS cooldown was assumed to be continued after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> until the RHR system in service temperature of 350'F is. reached.

Depressurization of the ruptured steam generator was then assumed to be performed immediately following the completion of the RCS cooldown. The ruptured steam generator was assumed to be

depressurized to the RHR in service pressure of 375 psia via steam release from the ruptured steam generator atmospheric steam dump valve, since this maximizes the steam release from the ruptured steam generator to the atmosphere which is conservative for the evaluation of the offsite radiation doses. The RCS pressure is also assumed to be reduced concurrently as the ruptured steam generator is depressurized. it is assumed that the continuation of the RCS cooldown.and depressurization to RHR operating conditions are completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident since there is ample time to complete the operations during this time period. The steam releases and feedwater f1qvs from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> were determined for the intact steam generator from s

-s,The steam released frorr. the ruptured steam generator from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was determined based on 0644D:10/071890 47

~

l

~

I

_ . i

0. , t.

After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, it is assumed that further plant cooldown to cold shutdown as well as long term cooling is provided by the RHR system. Therefore, the steam releases to the atmosphere are terminated after RHR in service conditions are assumed to be reached at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,

b. Mass Release Results The mass release calculations were performed using the methodology discussed above. For tae time period from initiation of the accident until leakage termination, the releases were determined from the LOFTTR2 results for the time prior to reactor trip and following reactor trip. Since the condenser is in service until reactor trip, any radioactivity released to the atmosphere prior to reactor trip will be through the air ejector discharge. After reactor trip, the releases to the atmosphere are assumed to be via the steam generator atmospheric steam dump valves. The mass release rates to the atmosphere from the LOFTTR2 analysis are l presented in Figures !!!.10 and !!!.11 for the ruptured and intact steam generators, respectively, for the time period until leakage termination.

l The mass releases calculated from the time of leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and from 2-8 hours are also assumed to be released to the atmosphere via the steam generator atmospheric steam dump valves. The mass releases for the SGTR event for each of the time intervals considered are presented in Table !!!.2. The mass releases prior to break flow termination, from break flow termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are summarized in Table !!!.3. The results indicate that approximately 43,500 lbm of steam are released from the ruptured steam generator to-the 0644D:1D/071890 48

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

i i

atmosphere in the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. A' total of 156,600 lbm of primary water is transferred to the secondary side of the ruptured

steam generator before the break flow is terminated..

~

i I

r 1

l 1

06440:10/071890 49 t'eM%@*"$1s9*T*s^--'-TerrvMt trt

• Mt-t' iw'Mi"++M*TTNWT*We M+"We W"W1==f $w 't d D' --YT%'&9-- Wu--tet'NMe9 eTr= w-f@MP" --F-9%%*-2 x e '-"M d 4+v t M=p 2-at'9 WhWut'tTre-tr*1r temWWW"m9r1es--t- 4 fmm 49-%.'eg-

l TABLE !!!.2 BVPS UNIT 2 SGTR ANALYSIS MASS RELEASES 0FFSITE RADIATION DOSE ANALYSIS TOTALMASSFLOW(POVNDS)

T!ME PERIOD 0 TRIP TRIP - TMSEP - TTBRK - T2 HRS -

TMSEP TTBRK T2 HRS TRHR Ruptured SG 1

- Condenser 177,400 0 0 0 0

- Atmosphere 0 42,200 1,300 0 36,000

- Feedwater 165,000 32,400 0 0 0 Intact SG Condenser 350,800 0 0 0 0

- Atmosphere 0 68,700 30,800 286,900 726,700

- Feedwater 350,800 177,200 75,100 310,200 735,200 Break Flow 12,400 115,800 28,400 0 0 TRIP = Time of reactor trip = 159 sec.

TMSEP = Time when water reaches the moisture separators = 2097 sec.

TTBRK = Time when break flow is terminated 3070 sec.

T2 HRS - Time at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> = 7200 sec.

TRHR = Time to reach RHR in service conditions, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> = 28,800 sec.

06440:10/071890 50 p -. .7- .--,..--.,..w. , - - , _ , . , . - - ,,w --

TABLE !!!.3 <

BVPS UNIT 2 SGTR ANALYSl$

SUMMARIZED MASS RELEASES OFFSITI RADIATION COSE ANALYSIS TOTAL MASS FLOW (P0llNO3) 0- TTBRK - 2 HRS -

TTBRK 2 HRS , 8 HRS Ruptured SG '

Condenser 177,400 0 0

- Atmosphere 43,500 0 36,000 Feedwater 197,400 0 0 Intact SGs Condenser 350,800 0 0 Atmosphere 99,500 286,900 726,700 Feedwater 603.100 310.200 735,200 Break Flow 156,600 0 0 06440:10/071890 S1 i

l

BEAVER VALLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE RUPTURED $0 ATMOSPHERIC MASS RELEASES ste..

s tee.'

sle..

[ tee.- i r

$250.-

E 2et..

t

" 150.

8

= ice. <

E

) 50. . N

' 'e . See, itee, tsee. tece. 25ee. seet. Esce.

TIMC iSCO)

Figure !!!.10 Ruptured $G Mass Release Rate to the Atmosphere -

Offsite Radiation Dose Analysis 06440:10/071890 52

I i

BEcvER vcLLEY UNIT 2 STEAM GENERATOR TUBE RUPTURE INTACT SOS ATMOSPHERIC NASS RELEASE 1400.<

&1200.

5

$1000.-

2 C

O 600.

E

?

u y 600.

0 E

c e 400.

8 G

200.

t

'O . 500, test. 1500. 2000. 2500. 5000. 5500.

TIMC ISCC)

Figure !!!.11 Intact SGs Mass Release Rate to the Atmosphere -

Offsite Radiation Oose Analysis r

a 06440:10/071890 53

1. 'T B. Offsite Radiation Dose Analysis The evaluation of the radiological consequences of a steam generator tube rupture, assumes that the reactor has been operating at the Technical Specification limit for primary coolant activity and primary to secondary ,

leakage for sufficient time to establish equilibrium concentrations of-  !

radionuclides in the reactor coolant and in the secondary coolant.

Radionuclides from the primary coolant enter the steam generator, via the ruptured tube, and are released to the atmosphere through the steam generator atmospheric steam dump valves (and safety valves) and via the air ejector discharge.

The quantity of radioactivity released to the environment, due to a SGTR, depends upon primary and secondary coolant activity, iodine spiking effects, primary to secondary break flow, break flow flashing fractions, attenuation of iodine carried by the flashed portion of the break flow, partitioning of iodine between the liquid and steam phases, the mass of fluid released from the generator and liquid-vapor partitioning in the turbine condenser hot well. All of these parameters were conservatively evaluated for a design basis double ended rupture of a single tube, 1

Desion Basis Analvtical Assumotions The major assumptions ahd parameters used in the analysis are itemized in Table III,4.

2. Source Term Calculations The radionuclide congentrations in the BVPS Unit 2 primary and secondary system, prior to and following the SGTR are determined as follows:

a.

The iodine concentrations in the reactor coolant will be based upon pre-accident and accident initiated iodine spikes.

0644D:10/101790 54 i

l

1. Accident Initiated Spike - The initial primary coolant iodine 1

concentration is I pC1/gm of Dose Equivalent (D.E.) 1-131.

Following the primary system depressurization associated with the SGTR, an iodine spike is initiated in the primary system which increases the iodine release rate from the fuel to the coolant to a value 500 times greater than the release rate that corresponds to the initial primary system iodine concentration. The duration of the spike is 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, ii. Pre-Accident Spike - A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine concentration from 1 to 60 pCi/ gram of D.E. I-131,

b. The initial secondary coolant iodine concentration is 0.1 pCi/ gram of D.E. 1-131.

c. The chemical form of iodine in the primary and secondary coolant is assumed to be elemental,

d. The initial noble gas concentrations in the reactor coolant are based on approximately 0.26't. fuel defects.

3. Dose Calculations The iodine transport model utilized in this analysis was proposed by Postma and Tam (Reference 4). The model considers break flow flashing, droplet size, bubble scrubbing, steaming, and partitioning.

The model assumes that a fraction of the iodine carried by the break flow becomes airborne immediately due to flashing and atomization. .

Removal credit is taken for scrubbing of iodine contained in the atomized coolant droplets as a function of the height of the secondary water level above the rupture site. The fraction of primary coolant iodine which is not assumed to become-airborne immediately mixes with L 06440:10/102090 55

the secondary water and is assumed to become airborne at a rate proportional to the steaming rate and the iodine partition coefficient. This analysis conservatively assumes an iodine partition coefficier.t of 0.01 between the steam generator liquid and steam phases. Droplet removal by the dryers is conservatively assumed to be negligible. The lodine transport :r.odel is illustrated in Figure 111.12.

The following assumptions and parameters were used to calculate the activity released to the atmosphere and the offsite doses following a SGTR.

a. The mass of reactor coolant discharged into the secondary system through the rupture and the mass of steam released from the ruptured and intact steam generators to the atmosphere are presented in Table III.2.

b. The time dependent fraction of rupture flow that flashes to steam and is immediately released to the environment is presented in Figure III.13. The break flow flashing fraction was conservatively calculated assuming that 100 percent of the break flow comes from the hot leg side of the steam generator, whereas the break flow actually comes from both the hot leg and cold leg sides of the steam generator,

c. In the lodine transport model, the time dependent iodine removal efficiency for scrubbing of steam bubbles as they rise from the ,

rupture site to the water surface conservatively assumes that the rupture is located at the intersection of the outer tube row n.nd the upper anti-vibration bar (approximately 4 inches below the apex of the tube bundle). However, the tube rupture break flow was conservatively calculated assuming that the break is at the top of the tube sheet. The water level relative to the top of the tubes in the ruptured and intact steam generators is shown in 06440:10/102090 56 l

l l

Figure !!!.14. The iodine scrubbing efficiency is determined by the method wggested by Postma and Tam (Ref. 4). The iodine scrubbing efficiencies are shown in Figure III.15.

The activity released to the environment by the flashed rupture flow can be written as follows:

A r

j IA II ~ 'II ) j J

where:

A r

- total iodine released to the environment by flashed primary coolant

= (integrated activity in rupture flow during time IA) intervalj) (flashing fraction for time interval j) eff)

= iodine scrubbing efficiency during time interval j

d. The total primary to secondary leak rate is assumed to be 1.0 gpm as allowed by the BVPS Unit 2 Technical Specifications. The leak rate is assumed to be 0.35 gpm for each of the intact steam generators and 0.3 gpm for the ruptured steam generator. The leakage to the intact steam generators-is assumed to persist for the duration of the accident.

e. The iodine partition factor between the liquid and steam of the ruptured and intact steam generators is assumed to be 0.01.

f. No credit was taken for radioactive decay during release and transport, or for cloud depletion by ground deposition during transport to the site boundary or outer boundary of the low population zone.

4 0644D:10/102090 57 l

g. Short.-term atmospheric dispersion factors (x/Qs) and breathing rates are provided in Table III.8. The breathing rates were obtained from NR'; Regulatory Guide 1.4. (Ref. 5).

4. Offtite Dose Calc.flation Modelt Offsite thyroid doses are calculated using the equation:

D

• OCI Th i (IAS)ij (ON)j (*/0)j i- j -

where (IAR)g) =

integrated activity of isotope i released during the time interval j in Cl*

(BR))

= breathing rate during time interval j in 3

meter /second (Table III.8)

(x/Q)3

. atmospheric dispersion factor during time interval j in second/ meter 3 (Table III.8)

(DCF)g . thyroid dose conversion factor via inhalation for isotope i in rem /Ci (Table 111.9)

D Th

. thyroid dose via inhalation in rem Offsite whole-body gamma doses are calculated using the equation:

D y 0.25{ l yg \j (IAR)q) (x/Q)3 1 /. ,

4 credit is taken for cloud depletion by ground deposition or by radioactive decay during transport to the exclusion area boundary or to the outer boundary of the low-population zone.

06440:10/101790 58

)

where:

(!AR);j =

!ntegrated activity of noble gas nuclide i re'ensed during time interval j in ci *

=

(xtQ)) atmospheric dispersion factor during time interval j in seconds /m3 i =

j average gama energy for noble gas nuclido i in Hev/ dis (Table 111,10)-

U y =

whole body gama dose due to imersion in rem Offsite beta skin doses are calculated using the equation:

O g = 0.23 {

.) - gg \j (lAR)4) (x/Q))/,

where:

=

(IAR)gj integrated activity of noble gas nuclide i released during time interval j in Ci *

=

(x/Q)) atmospheric dispersion factor during time interval j in seconds /m3 E gg -

average beta energy for noble gas nuclide i in Mev/ dis (Table 111.10)

Dg =

bata skin dose due to imersion in rem No credit is taken for cloud depletion by ground deposition or by radioactive decay during transport to the exclusion area boundary or to the outer boundary of the low population zone.

0644D:10/071890 59

_ _J

l

5. Resuits The calculated nuclide releases resulting from an SGTR are presented in Tablo III.11 for the pre-accident iodine spike case and in Table III.12 for the accident initiated iodine spike case. Thyroid, whole-body gamma, and beta-skin doses at the Exclusion Area Boundary and Low Population Zone are presented in Table 111.13. All doses are within the allowable guidelines as specified by Standard Review Plan 15.6.3 and 10CFR100.

t 06440:10/102090 60

l l

TABLE III 4 BVDS UNIT 2 SGIR_ ANALYSIS PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSEOUENCES, I. Source Data A. Core power level MHt 2766-B. ' Total steam generator tube 1.0  ;

leakage, prior to accident, gpm C. Reactor coolant iodine activity:

1. Accident Initiated Spike The initial RC iodine activities based.on 1 pCi/ gram of 0.E. I-131 are presented in

. Table III 5. The iodine appearance rates assumed for the accident initiated spike are presented in Table III 6.

2. Pre-Accident Spike Primary coolant iodine activities based on 60 pCi/ gram of D.E. I-131 are presented in Table III.5.

3. Noble Gas Activity The initial RC noble

. gas activities based on=0.261. fuel defects are presented in-Table'III.7.-

l

( 06440:10/102090 61-I L.. __

-l TABLE III.4 (Sheet 2)

D. Secondar:J system initial activity Dose equivalent of 0.1 pC1/gm of I-131

, presented in Table III.S.

1.91 x 10 8

~

E. Reactor coolant mass, grams 7

F. Initial steam generator water mass 4.5 x 10 (each), grams.

G. OffsiO Nur Lost at time of reattor trip H. Primary-to-secondary leakage 8 duration for intact SG, hrs.

I. Species of iodine 100 percent elemental II. Activity Release Data-A. Ruptured steam generator

1. Rupture flow See Table III.2

2. Rupture flow flashing fraction See-Figure III.13

3. Iodine scrubbing efficiency See Figure III.15-

4. Total steam release, lbs See: Table III.2

5. Iodine partition factor 0.01 1

+

0644D:1D/102090 62

-TABLE III.4 (Sheet 3) ,

6. Location of tube rupture Intersection of outer ,

-tube row and upper 4 anti-vibration bar- ,

8. Intact steam generators

1. Total primary-to-secondary 0.7 leakage, gpm

2. Total steam releaser Ibs. See Table III.2

3. Iodine partition factor. 0.01 C. Condenser

1. Iodine partition factor 0.01 D. Atmospheric Dispersion Factors .See Table III.8 06440:10/102090 63

i i

l TABLE III.5 BVPS UNIT 2 SGTR ANALYSIS IODINE SPECIFIC ACTIVITIES:

IE THE PRIMARY AND SECONDARY COOLANT BASED ON 1. 60 AND 0.1 uCi/aram 0F D.E.1-131 Soecific Activity (Iuci/am)

Primary Coolant Secondarv Coolant Nuclide 1 uC1/am 60 uCi/cm 0.1 uti/am I-131 0.66 39.9 0.069 I-132 0.23 13.9 0.020 I-133 1.0 .62.2 0.098 ,

I-134 0.14 ' 8.7 0.00045 I-135 0.55 33.4 0.044 l

06440:10/101790 64

I

-TABLE.III 6 BVPS UNIT 2 SGTR ANALYSIS IODINE SPIKE APPEARANCE RATES L (CURIES /SECOND)- I Eul I-132 I-133 I-134 1-135 i 1.36 2.52- 3.08 3.68- 2.81 i

i 1

L i

l 06440:1D/101790 .65-

TABLE III.7 BVPS UNIT-2-SGTR ANALYSIS NOBLE GAS SPECIFIC ACTIVITIES IN THE REACTOR COOLANT BASED ON 0.26% FUEL DEFECTS Sunliga Soncific Activity (uct /am)

Kr-83m 0.11 Kr-85m 0.55 Kr-85 2.90 Kr-87 0.32 Kr-88 0.84 Kr-89 0.027 Xe-131m 0.028-Xe-133m 0.81 Xe-133 6.9 ,

Xe-135m 0.29 Xe-135 0.85 Xe-137 0.043 Xe-138 0.18 r

l 0644D:10/102090 66 l

l

TABLE III.8 BVPS UNIT 2 SGTR ANALYSIS ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES limg Exclusion Area Boundary Low Population-3 Breathing (hours) 3 x/Q (Sec/m ) Zone x/Q (Sec/m ) Rate (;a3 /Sec) [5]

02 1.44 x 10'3' 7.07 x 10'3 3.47 x 10'4 28 -

7.07 x 10-5 3.47 x 10'4 l

06440:10/071890 67

' TABLE III.9 l BVPS UNIT 2 SGTR ANALYSIS THYR 0ID DOSE CONVERSION' FACTORS:-

(Rem / Curie) (Ref. 6).

i Nuclide I-131 1.48 x'10 6 4

I-132 5.35 x 10 I 133 l4.0 x 105 4 '

I-134 2.5 x 10 I 135 1.24 x.10 5 i

l h

06440:10/071890 68

y-..______.__..._ _ . _ _ . - . . . . _ . _ _ ._._ _ ._... _ . . _ . . _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .

I i

l TABLE III.10 BVPS UNIT 2 SGTR ANALYSIS AVERAGE GAMMA AND BETA ENERGY FOR NOBLE GASES (Hev/ dis) (Ref. 7T o i Nuclide Ey 50  !

i Kr-83m 0.0005 0.042 Kr-85m 0.156 0.253 Kr-85 0.0023 0.251 I Kr-87 0.793 1.33

, Kr-88 2.21 0.248 Kr-89 2.1 1.2  !

Xe-131m 0.0029 0.165 Xe-133m 0.02- 0.212 Xe-133 0.03 0.153 Xe-135m 0,43 0.099-Xe-135 0.246 0.325 Xe-137 0.19 1.8-Xe-138 1.2 . 0. 66--

0644D:10/102490 69

, l TABLE III.11 BVPS UNIT 2 SGTR ANALYSIS ENVIRONME G L RELEASES FOR PRE-ACCIDENT IODINE SPIKE CASE:

Total __ Releases (Ci)

Nuclide 0-2'hr 0-8 hr l

Kr-83m- 5.7 5.7 Kr-85m 3.1El -3.1 El Kr-85 1.7E2' 1,7E2 Kr-87 1.6El 11. 6 El ~

Kr-88 4.6El 4.6El Kr-89 1.7E-1 -1.7E-1 Xe-131m 1.7 L1,7 -

Xe-133m 4. 8 E1. 4.8E1 Xe-133 4 1E2 4.1 EE -

Xe-135m 8.2 8.2 Xe-135 4.9El 4.9El Xe-137 3.2E-1 3.5E-1 Xe-138 5.4. 5.4 im I-131 6. 5 E1. :6.6El I-132 - 2.1 E1: 12.1El I-133- -1.0E2. 1.0E2 I-134 1 ~. 2 E1 1.2El-I-135 5.3El 5.4E1 i

06440:10/102590 70

TABLE III.12

-BVPS UNIT 2 SGTR' ANALYSIS ENVIRONMENTAL-RELEASES FOR ACCIDENT INITIATED IODINE SPIKE CASE Total Releases (Ci)

Nuclide 0-2 hr 0-8 hr Kr-83m 5.7- 5.7 Kr-85m- 3.1El 3.1E1 Kr-85 1.7E2 .1.7E2 Kr-87 1.6El 1.6El Kr-88 4.6El- 4.6El Kr-89 1.7E-1 '1.7E-1 Xe-131m 1.7 1.7 Xe-133m 4.8E1 4.8E1 Xe-133 4.1E2- 4.1E2 Xe-135m 8.2- 8.2 Xe-135 4.9El :4.9El Xe-137 3.2E-1 3.5E-1 Xe-138 5.4 5.4 I-131 9.9- 1.1 El '

I-132 1.6El 1.7El I-133 2.2El 2.4El I-134 2.2E1 2.2El-I-135 1.9El 2.1El l

l 06440:10/102590 71 l

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

i TABLE III.13 y l

BVPS UNIT 2 SOTR ANALYSIS' OFFSI1E RADIATION DOSES 1

Doses (Rem)

Calculated- Allowable

-Value Guideline Value [Ref 81' '

l. becident Initiated Iodine Soike Exclusion Area Boundary (0-2 hr.)

Thyroid 13.4- 30 Whole-Body Gamma 0.2 :2.5 Beta-Skin 0.2 2.5 Low Population Zone (0-8 hr.)

Thyroid 0.8 30 Whole-Body Gamma 0.009 2.5 Beta-Skin 0.007 2.5

2. Pre-Accident Iodine Soike Exclusion Area Boundary (0-2 hr.)

Thyroid 71.6 300*

Whole-Body Gamma 10 . 2 25*-

Beta-Skin 0.1 25*

Low-Population Zone (0-8-hr.)

Thyroid 3.6 300*

Hhole-Body Gamma .0.007 25*

8 eta-Skin 0.005 25*

• Doses should be appropriately within the guideline values.

06440:10/102590 72

m

? SCRUBBING >

STEAM PRIMARY SHING + VOLUME _ >

C00LAN ATMOSPHERE LIQUID k U PARTITION SECONDARY COOLANT Figure III.12 Iodine Transport Model - Offsite Radiation Oose Analysis 06440:10/101790 73

r BEAVER VALLEY UNIT 2 STEAf1 GENERATOR TUBE RUPTURE BREAK FLOW FloSHING. FRACTION

.16 9

.16 .

g .14

$.12 5

= .1 0

2 k.00-y .06-E

.e4

.02<

\

'O. 500. ~

1998. 1500. 2000. 2500, 5000, 7520, T!NC 1 SCC)

Figure-'III.13 Break Flow Flashing Fraction - Offsite Radiation-Dose Analysis I

i 1

06440:10/101790 74

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

l l

i BEAVER VALLEY UNIT 2 STEAf1 OENERATOR TUBE RUPTURE 1 SO SECONDARY LEVEL ABOVE TOP OF TUBES 225.

I 200. "

$ RUPTURED SG m 175.

W

= i5e.

b g 125.

W 100, ia 75.<

W d 50. %

5 INTACT SGs 25.

v d O.-

8 25.

50 8. 588. 1908. 1589. 2944.- 2588. 5990. 5500.

TIME (SCC) i Figure.III.14 SG' Hater Level Above Top of Tubes -

'Offsite Radiation Dose Analysis-06440:10/101790 75 1

w t+y--1

0.06 BERVER VALLEY UNIT 2 SGTR 0.05 -

M g 0.04 -

o E

0.03 - -

0.02 -

m 0.01 0.00 -

0 250 .500 750- 1000 1250 1500 TIME (SEC0tOS)-

l

[

I Figure III.15 Iodine Scrubbing Efficiency'- Offsite Radiation-Dose Analysis 06440:10/101790 76 l

IV. CONCLUSION An evaluation has been performed for a design basis SGTR for Beaver Val Power Station Unit 2 to demonstrate that the potential consequences are acceptable.

An analysis was performed to demonstrate margin to steam generator overfill with the limiting single failure relative to overfill. The limiting single failure is the failure of the

~

~

iheresultsof this analysis indicate that the recovery actions can be completed to terminate the primary to secondary break flow before overfill of the ruptured steam

' generator would occur.

Since it is concluded that steam generator overfill will not occur for a design basis SGTR, an analysis was also performed to determine the offsite radiation doses assuming the limiting: single failure for offsite doses. For this analysis, it was assumed that the ruptured steam generator atmospheric steam dump valve fails open at the time the ruptured steam generator is isolated, and that the failed open valve must be isolated by locally closin the associated block valve. The primary to secondary break flow and the mass releases to the atmosphere were determined for this case, and the offsite radiation doses were calculated using this information. The resulting doses at the exclusion area boundary and low population zone are within the allowable guidelines as specified by Standard Review Plan 15.6.3 and 10CFR100.

Thus, it is concluded that the consequences of a design basis steam generator tube rupture at Beaver Valley Power Station Unit 2 would be acceptable.

l l

l l

06440:10/101790 77

V. REFERENCES

1. Lewis, Huang, Behnke, Fittante, Gelman, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," HCAP-10698-P-A (PROPRIETARY)/WCAP-10750-A [NON-PROPRIETARY). August 1987.

2. Lewis, Nuang, Rubin, " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident," Supplement I to HCAP-10698-P-A (PROPRIETARY]/ Supplement 1 to HCAP-10750-A (NON-PROPRIETARY), March 1986.

3. Lewis, Huang, Rubin, Murray, Roidt, Hopkins, " Evaluation of Steam Generator Overfill Due to a Steam Generator Tube Rupture Accident,"

HCAP-11002 (PROPRIETARY]/HCAP-11003 (NON-PROPRIETARY), February 1986.

4. Postma, A. K., Tam, P. S., " Iodine Behavior in a PHR Cooling System Following a Postulated Steam Generator Tube Rupture", NUREG-0409,

5. NRC Regulatory Guide 1.4, Rev. 2. " Assumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for Pressurized Hater Reactors", June 1974.

6. DiNunno, J. J., et, al., " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, March 23, 1962.

7. Bell, H. J., "0RIGEN - The ORNL Isotope Generation and Depletion Code,"

l ORNL-4628, 1973.

8. Standard Review Plan, Section 15.6-3, " Radiological Consequences of Steam

. Generator Tube Failure," NUREG-0800, July 1981.

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