ML20147D328

LOFTTR2 Analysis for Steam Generator Tube Rupture Event. Related Info Encl

ML20147D328
Person / Time
Site:	Vogtle
Issue date:	01/31/1988
From:	Robert Lewis, Mendler O, Miller T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19341E001	List: ML20147D233 ML20147D277 ML20147D314 ML20147D328
References
WCAP-11732, NUDOCS 8803030289
Download: ML20147D328 (84)

Text

WESTINGHOUSE CLASS 3 WCAP 11732 I

i LOFTTR2 ANALYSIS FOR A STEAN GENERATOR TU8E RUPTURE EVENT FOR THE V0GTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 R. N. Lewis

0. J. Mendler T. A. Miller

' K. Rubin JANUARY 1988 Nuclear Safety Department P

1 Westinghouss Electric Corporation Nuclear Energy Systems '

P.O. Box 355

1. Pittsburgh, Pennsylvania 15230 I,) 1988 by L'estinghouse Electric Corporation  ;

mw o===. "BR 9 88R 82888 a P PDR

TABLE OF CONTENTS ,

Pages I. INTRODUCTION 1

!!. ANALYSIS OF MARGIN TO STEAN GENERATOR OVERFILL 4 A. Design Basis Accident 4

8. Conservttive Assumptions 5 C. Operator Action Times 7

0. Transient Description 13  !

III. ANALYSIS OF OFFSITE RADIOLOGICAL CONSEQUENCES 26 A. Thermal and Hydraulic Analysis 26

1. Design Basis Accident 26 s

2. Conservative Assumptions 27

3. Operator Action Times

, 29

4. Transient Description 29

. 5. Ness Releases 44 t

B. Offsite Radiation Dese Analysis 53 IV. CONCLUSION 70 V. REFERENCES 71 P

O I-it7s<:to n n oi n i j

LIST OF TABLES Table Title Page 11.1 Operator Action Times for Design Basis Analysis 12  :

!!.2 Sequence of Events - Wargin to Overfill Analysis 18

!!!.1 Sequence of Events - Offsite Radiation Dose Analysis 34

!!!.2 Mass Releases - Offsite Radiation Oose Analysis 49

!!I.3 Summarized Mass Releases - Offsite Radiation Dose 50 Analysis III.4 Parameters Used in Evaluating Radiological 58 Consequences

!!!.5 lodine Specific Activities in the Primary and Secondary 61 Coolant

!!!.6 lodine Spike Appearance Rates 62

!!!.7 Atmospheric Dispersion Factors and Breathing Rates 63 i

!!!.8 Thyroid Dese Conversion Factors 64

!!!.9 Offsite Radiation Ooses 65 J

1 1179vilo/c20164 ii i

LIST OF FIGURES Floure Title M

!!.1 Pressurizer Level - Margin to Overfill Analysis 19

!!.2 RCS Pressure - Margin to Overfill Analysis 20

!!.3 Secondary Pressure - Margin to Overfill Analysis 21 II.4 Intact Loop Hot and Cold Leg RCS Temperatures - 22 Margin to Overfill Analysis

!!.5 Reactor Coolant Average Temperature - Margin to 23 Overfill Analysis II.6 Primary to Secondary Break Flow Rate - Margin to 24 >

Overfill Analysis

!!.7 Ruptured SG Water Volume - Margin to Overfill Analysis 25

!!!.1 RCS Pressure - Offsite Radiation Dose Analysis 35 l

l  !!!.2 Secondary Pressure - Offsite Radiation Dese Analysis 36 f

!!!.3 Pressurizer Level - Offsite Radiation Dese Analysis 37

!!!.4 Ruptured Loop Hot and Cold Leg RCS Temperatures - 38 Offsite Radiation Dese Analysis

!!!.5 Intact Loop Hot and Cold leg RCS Temperatures - 39 Offsite Radiation Dese Analysis t

,  !!!.6 Differential Pressure Between RCS and Ruptured 40 SG - Offsite Radiation Dose Analysis -

l ti7kionaoisa 111

i

!.!STOFFIGURES(Cont) {

Figure Title M

!!!.7 Primary to Secondary Break Flow Rate - Offsite 41 Radiation Dose Analysis

!!!.8 Ruptured SG Water Volume - Offsite Radiation Dose 42 Analysis

!!!.9 Ruptured SG Water Wass - Offsite Radiation Oose Analysis 43

!!!.10 Ruptured SG Mass Release Rate to the Atmosphere - 51 Offsite Radiation Doss Analysis i

!!!.11 Intact SGs Mass Release Rate to the Atmosphere - 52 Offsite Radiation Dose Analysis

!!!.12 lodine Transport Model - Offsite Radiation Dose Analysis -

66

!!!.13 Break Flow Flashing Fraction Offsite Radiation 67 Dose Analysis

!!!.14 SG Water Level Above Top of Tubes - Offsite 68 Radiation Oose Analysis

!!!.15 lodine Scrubbing Efficiency - Offsite Radiation Dese 69 Analysis -

1179tle/C201sa iV t

1. INTRODUCTION An evaluation for a design basis steam generator tube rupture (SGTR) event has been performed for the Vogtle Electric Generating Plant (Plant Vogtle), Units 1 and 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 desee are less than the allowable guidelines.

Plant Vogtle employs two essentially identical Westinghouse pressurized water reactor (PWR) units rated at 3411 mwt. The reactor coolant system for each unit has four reactor coolant loops with Model F steam generators. Since the reactors, structures, ard all auxiliary equipment are substantially identical for the two units, the SGTR evaluation is applicable for both units. It is noted that Unit 1 is currently licensed to operate with Westinghouse standard fuel with a negative moderator temperature coefficient. However, it is anticipated that the Technical Specifications will be changed to permit eperation with a positive moderator temperature coefficient for future fuel cycles. Therefore, the more limiting parameters for operation with a positive moderator temperature coefficient were used for the SGTR evaluation such that the results are applicable for the current licensing basis as well as for ccaratien with a positive moderator temperature coefficient.

The steam generator tube rupture analyses were performed for Plant Vogtle using the methodology developed in WCAP-10750 (Reference 1) and Supplement 1 i

to WCAP-10750 (Reference 2). *his analysis methodology was developed by the SGTR Subgroup of the Westinghouse Owners Group and was approved by the NRC in Safety Evaluation Reports dated December 17, 1985 and Warch 30, 1987. Plant

response to the event was modeled using the LOFTTR2 computer code with conservative assumptions of break size and location, condenser availability and initial secondary water mass in the ruptured steam generator. The enalysis methodology includes the siMation of the operator actions for i recovery from a steam generator tubs et"re based on the Plant Vogtle I

. E ergency Operating Procedures, whir.b were developed from the Westinghouse F

1179v.to/01284a 1

OwnersGroupEmergencyResponseGuidelines(ERGS). In subsequent references to the Plant Vogtle Emergency Operating Procedures, the specific Yogtle , ,

Emergency Operating Procedure will be listed along with the corresponding Westinghouse Owners Group ERG in parenthesis. ,

Since the limiting single failure is different for the overfill analysis and the offsite radiation dose analysis, the two analyses were performed using different single failure assumptions. For the margin to overfill analysis, it was assumed that the

~

~hertheanalysisoftheoffsiteradiationdoses, ~

]1heseassumptionsare i consistent with the methodology in References 1 and 2.

The LOFTTR2 analysis to determine the margin to overfill was performed for the time period from the tube rupture until the primary and secondary pressures are equalized and the break flow is terminated. The water volume in the

secondary side of the ruptured steam generator was calculated as a function of -

time to demonstrate that overfill does not occur. This analysis demonstrates that there is cargin to steam generator overfill for Plant Vogtle.

For the offsite radiation dose analysis, the primary to secondary break flow and the steam releases to the atmosphere from both the ruptured and intact steam gcr.arators 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 l l time period following break flow termination, steam releases from and l 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 I was used to calculate the radiation doses at the exclusion area boundary and *(

- low pepulation zone assuming that the prirsary coolant activity is at the l

l l

1 m s a w s: 2 l

i reximum allowable Technical Specification limit prior to the accident. This j

. analysis demonstrates that the offsite doses for Plant Vogtle are well within  !

the allowable guidelines specified in the Standard Review Plan, NUREG-0800, .

- Section 15.6.3, and 10CFR100.  ;

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

!!. ANALYSIS OF MARGIN TO STEAN GENERATOR OVERFILL An analysis was performed to determine thi margin to steam generator overfill for a design basis SGTR event for Plant Vogtle. The analysis was performed using the LOFTTR2 program and the methodology developed in Reference 1. 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. Design Basis Accident The accident modeled is a double-ended break of one steam generator tube

! f the tube sheet locatedatthetopA'halocationofthebreak l

] _

-eL ftwas also assumed that loss of offsite power occurs at the time of reactor trip, and the worst rod was assumed to be stuck at the operating position at the time of reactor trip.

The most limiting single failure with respect to steam generator overfill was determined to be _

,do, wever, a sensitivity l study indicates that the most limiting single failure for the four-loop Plant Vogtle units is

~

l N Plant Vogtle AFW system consists of two motor-driven pumps, and one turbine-driven pump with a capacity equal to approximately the combined capacity of the two motor-driven pumps. Each

! motor-driven pump normally feeds two staae generators and the turbine-driven pump feeds all four steam generators. There are two AFW flow control valves for each steam generator, one in the flow path fron the motor-driven pump and one in the flow path from the turbine-driven -

pump. The AFW flow control valves are normally open and are used to terminate feedwater flow to the ruptured steam generator and control '

l l

117sc10/0203 a 4'

inventory in the intac,t steam generators. w 6,4, requires the  !

• ~ ~

operator to perform additional actions to I

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- en .

It was assumed that  ;

the  ;

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InaccordancewithReference1,it

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was assumed that I

- as e,s This results in >

additional primary to secondary leakage as well as

~

! ~ldch

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decreases the margin to steam generator overfill.

)

B. Conservative Assumptions  !

. Sensitivity studies were performed previously to identify the initial j plant conditions and analysis assumptions which are conservative relative to steam generator overfill, and the results of these studies were  ;

reported in Reference 1. The conservative conditions and assumptions t which were used in Reference 1 were also used in the LOFTTR2 analysis to determine the margin to steam generator overfill for Plant Vogtle with the  !

exception of the following differences.  ;

1. Reactor Trip and Turbine Runback f i

A turbine runback can either be initiated automatically or the j

[ -

operator can manually reduce the turbine load following an SGTR to

.atteset to prevent a reactor trip. For the reference plant analyst l

- in WCAP-10/50 reactor trip was calculated to occur at approximately  !

- gand4turbine runback to a.c -

, was

~

steulated based on a runback rate of N effect of l 1

i i

mw.wunn 5 l

l

turbine runback was conservatively simulated by l

~#C .i Howevor, if reactor trip

, e, c l occurs prior to ,'urbine t runback to

~ ~

~

would not be po:Tible. 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  !

1 AFW 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 after the SGTR occurs ,  ;

rather than at a fixed time after reactor trip._ It woul_dg overly l conservative to include the turbine runback to in addition to the penalty in secolidary mass due to earlier reactor trip. Thus, for this analysis, the time of reactor trip was determined by modeling tra Plant Vogtle reactor protection system, and turbine runback was , 4,c

.imulated

2. Steam Generator Secondary Mass 8, d A initial secondary water mass in the ruptured steam generator .

~ ~

was determined by Reference 1 to be conservative for overfill. As noted above, turbine runbar.k, was assumed to be initiated and was i,

simula

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by he initial steam generetnr total fluid mass was conservatively

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l 3. AFW System Ooeratien l For_the reference plant analysis in WCAP-10750, reactor tgig occurred j' on after the ,g,c,

. SGTR, and 51 was initiated on low pressurizer pressure at

~

after reacter trip. The reactor and turbine trip and the assumed concurrent loss of offsite power will result in the termination of 1

4 1179v 10/0128ta fa

l l

main feedwater flow and actuation of the AFW system. The SI signal will also result in automatic isolation of the main feedwater system  ;

and actuation of the AFW system. The flow from the turbine - driven i

~

AFW pump will be available within approximately 10 seconds following the actuation signal, but the flow from the motor-driven AFW pumps will not be available until approximately 60 seconds due to the startup and load sequencing for the emergency diesel generators. For the reference plant analysis, it was assumed that AFW flow from both the turbine and motor-driven pumps [s initiated The total AFW flow from all of the AFWpumpswasassumedtobedistrIbuteduniformlytoeachofthesteam generators until operator actions are simulated to throttle AFW flow to control steam generator water level in accordance with the emergency procedures.

It is noted that if reactor trip occurs on

- - n, C the pressure at

~

the time of reactor trip may be significantly higher than the SI

. initiation setpoint. In this event, there may be a significant time delay between reactor trip and SI initiation, and it would not be conservativetomedgelthe, Thus, for this analysis, the time of reactor trip

~

was determined by codeling the Plant Vogtle reactor protection system, and thegetuation of the AFW system was based on the ~

It was assumed that flow from both the turbine and motor-driven AFW pumps is initiated at the

- a, c The total AFW flow assumed for the analysis is the combined capability of the turbine-driven pump and both motor-driven pumps C. 0::erator 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 Plant Vogtle Emergency 1179<:1o/012888 7

Operating Procedure 19030-1 (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 depressurhation of the RCS to restore inventory, .and termination of SI to ~

stop priaary to secondary leakage. These operator actions are described below

1. Identify the ruptured steam generator.

High secondary side activity, as indicated by the main steamline radiation monitors, the condenser air ejector radiation monitor, or steam generator blowdown radiation monitors, typically will provide the first indication of an SGTR event. The ruptured steam generator can be identified by an unexpected increase in steam generator narrow range level or a high radiation indication on the corresponding main steamline radiation monitor. For an SGTR that results in a reactor trip at high power as assumed in this analysis, the steam generator water level will decrease to near the bottom of the narrow range for all of t.ne steam generators. The AFW flow will begin to refill the ,

steam generators, distributing approximately equal flow to each of the steam generators. Since primary to secondary leakage adds additional .

l liquid inventory to the ruptured steam generator, the water level will increase more rapidly in that steam generator. This response, as indicated 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 to minimizing radiological releases, this also reduces the possibility of overfilling the ruptured steam -

I generator with water by 1) minid zing the accumulation of feedwater flow and 2) enabling the operator to establish a pressure differential 1179v:10/012888 8

between the ruptured and intact steam generators as a necessary step toward' terminating primary to secondary leakage. In the Plant Vogtle Emergency Operating Procedure for steam generator tube rupture

)

recovery, the operator is directed to terminate the feed flow to the l ruptured steam generator if the level indication is greater than 5% on the narrow range instrument, and the steam generator level is noga}ly controlled to a maximum of 50%. it was assumed that the ruptured steam generator would be isolated when

- a, C Theapplicationofthisassumptiontg Plant Vogtle would result in the use of for isolation of the ruptured steam generator.

a, c.

Thus, for the Plant Vogtle analysis, the ruptured steam generator was assumed to be isolated at 33 percent narrow range level or at 12 minutes, whichever was longer.

3. Cool down the Reactor Coolant System (RCS) using the intact steam generators.

After isolation of the ruptured steam generator, the RCS is cooled as rapidly as possible to less 'han 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 PORVs on the intact steam generators. Since

. offsite power is assumed to be lost at reactor trip for this analysis, the cooldown was performed by dumping steam via the PORVs on the three intact steam generators.

1179v:1o/012ss: 9

4. Depressurize the RCS to restore reactor coolant inventory.

When the cooldown is completed, SI flow will increase RCS pressure until break flow matches SI flow. Consequently, SI flow must be terminated to stop primary to secondary 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 indication after SI flow is stopped. Since leakage from the primary side will continue after SI flow is stopped until RCS 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 when RCS pressure equals the pressure in the ruptured steam generator.

The RCS depressurization is performed using normal pressurizer spray if the reactor coolant pumps (RCPs) are running or the pressurizer PORVs if the RCPs are not running. 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. Therefore, for this analysis, RCS depressurization was performed using a pressurizer PORV. .

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 I

completed, SI flow must be stopped to terminate primary tc secondary

( leakage. Primary to secondary leakage will continue after SI flow is I

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.

l 1179<:to/oissa 10

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 explicitly modeled, it is also necessary to account for the' time required to perform the steps. It is noted that the total time required to complete 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 cperations (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 safety injection termination were developed in Reference 1 for the design basis analysis. Georgia Power Company has established the corresponding operator action times to perform these operations for the Plant Vogtle units as reported in Reference 3. The operator actions and the corresponding operator action times for Plant Vogtle are listed in Table 11.1.

O i

O 1179v.1o/0203Sa 11

TABLE II.1 PLANT V0GTLE SGTR ANALYSIS .

OPERATOR ACTION TIMES FOR DESIGN BASIS ANALYSIS Action Time (min)

Identify and isolate ruptured SG 12 min or LOFTTR2 calculated time to reach 33% narrow range level in the ruptured SG, whichever is longer Operator action time to initiate 7 cooldown Cooldown Calculated by LOFTTR2 Operator action time to initiate 2 depressurization Depressurization Calculated by LOFTTR2 ,

Operator action time to initiate 2 SI termination SI termination and pressure Calculated time for SI termination equalization and equalization of RCS and ruptured SG pressures 1179v:1o/012ssa 12

i D. Transient Description The LOFTTR2 analysis results for the margin to overfill analysis are described below. The sequence of events fce this transient is presented in Table 11.2.

Following the tube rupture, reactor coolant flows from the primary into the secondary side of the ruptured steam generator since the primary pressure is greater than the steam generator pressure. In response to this loss of reactor coolant, pressurizer level decreases as shown in Figure II.l. The RCS pressure also decreases as shown in Figure II.2 as the steam bubblo in the pressurizer expands. As the RCS pressure decreases due to the continued primary to secondary leakage, automatic reactor trip occurs on an overtemperature delta-T trip signal.

After reactor trip, core power rapidly decreases to decay heat levels.

The turbine 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 dump valves remain closed due to the loss of condenser vacuum resulting a

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  ;

PORVs (and safety valves if their setpoints are reached) lift to dissipate the energy, as shown in Figure II.3. The main feedwater flow will be terminated and AFW flow will be automatically initiated following reactor trip and the loss of offsite power.

The RCS pressure decreases more rapidly after reactor trip as energy transfer to the secondary shrinks the reactor coolant and the tube rupture break flow continues to deplete primary inventory. The decrease in RCS Inventory results in a low pressurizer pressure SI signal. Pressurizer level also decreases more rapidly following reactor trip. After SI actuation, the SI flow rate exceeds the tube rupture break flow rate and  ;

1179v:lo/012sta 13

1 the pressurizer level begins to increase. This also results in an increase-in the RCS pressure which trends toward the equilibrium value where the SI flow rate equals the break flow rate. -

It is noted that the pressurizer level increases to approximately 55% and -

the RCS pressure increases to approximately 2580 psia before the RCS cooldown is performed. Although this pressure is above the pressurizer PORY and safety valve setpoints, operation of the relief valves was not modeled in the analysis since this maximizes the RCS pressure which results in a conservative calculation of the primary to secondary break flow. The calculate 1 RCS pressure is also conservatively high since the analysis is based on conservative maximum SI flow rates which are l significantly higher than best estimate SI flow rates. With more j realistic SI flow rates, it is possible that the RCS pressure will remain I below the pressurizer PORY setpoint for a design basis SGTR. Although the l pressurizer PORY setpoint may not be reached for a design basis SGTR, it is expected that the setpoint would be reached for smaller ruptures if SI is actuated. However, operation of the pressurizer PORVs should not affect the operator response to the SGTR since the recovery procedure is designed to mitigate an SGTR even if this should occur. It is noted that .

the subsequent operator actions result in the RCS pressure being reduced to below the PORV setpoint, and ultimately to the ruptured steam generator ,

pressure to terminate the primary to secondary break flow.

i Since offsite power is assumed lost at reactor trip, the RCPs trip and a i gradual transition to natural circulation flow occurs. Innediately following reactor trip the temperature differential across the core decreases as core power decays (see Figure 11.4), however, the temperature differential subsequently increases as natural circulation flow develops.

The cold leg temperatures trend toward the steam generator temperature as i the fluid residence time in the tube region increases. The RCS temperatures continue to slowly decrease due to the continued AFW flow to the steam generators until operator actions are taken to control the AFW ,

flow to maintain the specified level. The reactor coolant average i temperature response is shown in Figure 11.5. .

117seio/caosas 14 l

Major Operator Actions

1. Identify and Isolate the Ruptured Steam Generator -

Once a tube rupture has been identified, recovery actions begin by isolating steam flow from the ruptured steam generator and throttling 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 33% on the ruptured steam generator or at 12 minutes after initiation of the SGTR, whichever is longer. For the Plant Vogtle analysis, the time to reach 33% is less than 12 minutes, and thus the ruptured steam generator is assumed to be isolated at 12 minutes.

As noted previously, the limiting single failure was assumed to

-acr

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when the

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isolation is performed. It was assumed that

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2. Cool down the RCS to Establish Subcooling Margin After isolation of the ruptured steam generator is completed at 14 minutes, a 7 minute operator action time is imposed prior to initiating the cooldown. The actual delay time used in the analysis is 2 seconds longer because of the computer program numerical requirements for simulating the operator actions. After this time, actions are taken to cool the RCS as rapidly as possible by dumping steam from the intact steam generators. Since offsite power is lost, theRCSiscooledbydumpingsteamtotheatmosphereusingtheP0Rfs _ ,,

on the intact steam generators. It was assumed that , _

tbe intact steam generator PORVs are opened at 1262 seconds for the RCS

, . cooldown. The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20'F plus an allowance of 24'F r

1179v:10/012ssa 15

for subcooling uncertainty. When these conditions are satisfied at 1842 seconds, it is assumed that the operator closes the intact steam .

generator PORVs to 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 II.3, and the effect of the cooldown on the RCS temperature is shown in Figures II.4 and 11.5.

The RCS pressure also decreases during this cooldown process due to shrinkage of the reactor coolant as shown in Figure II.2.

3. Depressurize RCS to Restore Inventory After the RCS cooldown, a 2 minute operator action time is included prior to depressurization. The RCS depressurization is initiated at 1964 seconds to assure adequate coolant invantory 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 depressurization is continued until ar.y of the following

~

conditions are satisfied: RCS pressure is less than the ruptured steam generator pressure and pressurizer level is greater than the ,

allowance of 9% for pressurizer level uncertainty, or pressurizer level is greater than 69%, or RCS subcooling is less than the 24*F allowance for subcooling uncertainty. The RCS depressurization reduces the break flow as shown in Figure II.6, and increases SI flow to refill the pressurizer as shown in Figure 11.1. Although the pressurizer level is less than 69% when the pressurizer PORV is closed, the level continues to increase to a maximum of approximately 77% when SI flow is terminated.

4. Terminate SI to Stop Primary to Secondary Leakage The previous actions have established adequate RCS subcooling, verified a secondary side heat sink, and restored the reactor coolant ,

inventory to ensure that SI flow is no longer needed. When these 1179v:10/020184 16 t

actions have been completed, the SI flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary i leakage. The SI flow is terminated at this time if RCS subcooling is greater than the 24'F allowance for subcooling uncertainty, minimum I AFW flow is available or at least one intact steam generator level is in the narrow range, the RCS pressure is increasing, and the pressurizer level is greater than the 9% allowance for uncertainty.

To assure that the RCS pres 3.ure is increasing, SI is not terminated in the analysis until the RCS' pressure increases by at least 50 psi.

After depressurization is completed, an operator action time of 2 minutes was assumed prior to SI termination. Since the above requirements are satisfied, SI termination was performed at this time. After SI termination, the RCS pressure begins to decrease as shown in Figure II.2. The intact steam generator PORVs also automatically open to dum steam to maintain the prescribed RCS temperature to ensure that subcooling is maintained. When the PORVs 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 primary to secondary leakage continues af ter the SI flow is terminated until the RCS and ruptured steam generator pressures equalize.

The primary to secondary break flow rate throughout the recovery operations is presented in Figure II.6. The water volume in the ruptured steam generator is presented as a function of time in Figure II.7. 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 5906 ft3 . Therefore, it is concluded that overfill of the ruptured steam generator will not occur for a design basis SGTR for Plant Vogtle.

O e

l tinnto/otosu 17

4 TABLE II.2 PLANT V0GTLE SGTR ANALYSIS SEQUENCE OF EVENTS MARGIN TO OVERFILL' ANALYSIS EVENT Time (sec)

SG Tube Rupture O Reactor Trip 114 SI Actuated 300 Ruptured SG Isolated 840 RCS Cooldown Initiated 1262 RCS Cooldown Terminated 1842

! RCS Depressurization Initiated 1964 i

RCS Depressurization Terminated 2068 l

SI Terminated 2188 l Steam Relief to Maintain RCS Subcooling 2956 Break Flow Terminated 3768 i

I l

l 1179v:10/020184 18

o V0GTLE STEAM GENERATOR TUBE RUPTURE

. MARGIN TO OVERFILL ANALYSIS PRESSURIZER LEVEL

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es.

75.

70.

65.

(

'$ 62.~\

s y55.

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& 50.

2 f45.

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, $40.

55.

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25.

'O . 500. 1223. 1522. 2220, 2520. 5000. 5520. 4000.

TIME ISCCI Figure 11.1 Pressurizer Level - Margin to Overfill Analysis 4

e 117sv.1o/o12:ss 19

V0GTLE STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS RCS PRESSURE 2622.-

2422.

2222.<\

3 -

f2222.

Y G 1922.

O C 1622. ,

u 1422.. .

1222.

1222 2. 522. 1222. 1522. 2222. 2522. 5222. 5522. 4222.

TIMC (SCC)

Figure !!.2 RCS Pressure - Margin to Overfill Analysis ,

1179v:10/012ssa 20

l V0GTLE STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS SECONDARY PRESSURE 1222.

RUPTURED SG 1122.

1222. V

.E.

E S22.

INTACT SGs 5 '

G E22.

C

> 722.

c 6f2. r

. U 522.

420.

r

2. 522. 1220. 1520. 2222. 2522. 5222. 5522. 4222.

TIME ISEC)

Figure !!.3 Secondary Pressure - Margin to Overfill Analysis i 1179v:le/012484 21

V0GTLE STE Af1 GENERATOR TUBE RUPTURE MARGIN - 70 OVERFILL ANALYSIS INTACT LOOP HOT AND COLO LEG RCS TEMPERATURES 650..

600. THOT C

$550.

Sec. TCOLD E

~

._ 452. -

9 y400. '

M 550, 2

52: .3. 520. 1222. 1522. 2200. 25C2, 5200. 5520. 4000.

TIME (SEC) ,

Figure 11.4 Intact Loop Hot and Cold Leg RCS Temperatures -

Margin to Overfill Analysis ,

It?s<.io/012ssa 22 i

V0GTLE STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS REACTOR COOLANT AVERAGE TEMPERATURE 650.

C e 600.

8_

g550.

i 1

U 500.

I e s 450.

a 400, s

>=

r-8 g 550.

500.O. 520. 1000. 1500. 2000. 2500. 5000. 5500. 4000.

71 r'C I SC C I Figure II.5 Reactor Coolant Average Temperature -

Margin to Overfill Analysis ,

i itnvio/otassa 23

.,.s._ .

V0GTLE STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS PRIMARY TO SECONDARY BREAK FLOW 50.

45.

N 40.

55.

C ss 50.

5

- 25.

I 5

u 20.

5

• t 5.

=

T t 10. .

l s.

l t

l

0. W 5 .O . 500. 1000. 1500. 2000. 2500. 5000. 5500. 4000.

! T!FC (SCC 1 l

l Figure !!.6 Primary to Secondary Break Flow Rate -

1 Margin to Overfill Analysis I

1 1

s in.:io<oian 24

k v0GTLE STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS RUPTURED SG WATER VOLUME saee. Model F SG Secondarg Volume _ _ _ , , _ _

5502.

G 5 5222.

d

> 4522.

3 3

3 r y 4:20.

8 h

, T 5500.

5:22.

25: .

O. 520. 1000. 1500, 2220. 2500. !220. 5500. 4200.

TIME ISEC) i i

Figure !!.7 Ruptured SG Water Volume - Margin to  ;

Overfill Analysis 1

117saozoisses 25

III. ANALYSIS OF 0FFSITE RADIOLOGICAL CONSEQUENCES An analysis was also performed to determine the offsite radiological consequences for a design basis SGTR event for Plant Vogtle Units 1 and 2. ,

The thermal and hydraulic and the offsite radiation dose analyses were performed using the methodology developed in References 1 and 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 ceindenser and to the atmosphere were calculated until break flow termination with the LOFTTR2 program for use in calculating the offsite radiation doses. This section provides a discussion of the methods and assum tions 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. Design Basis Accident The accident modeled is a double-ended break of one steam genera'c -

~

tube sheet -

tube located at the top ofi thg $e location 7f the break

~ ~

4 C.

l lt was also assumed that loss of offsite

power occurs at the time of reactor trip and the worst rod was assumed to be stuck at the operating position at the time of reactor trip.

l l Based on the information in Reference 2, the most limiting single failurewithrespecttoof{failureofsfte doses is which -

will increase primary to secondary leakage and the mass release to the atmosphere. Pressure in the ruptured steam generator wiil remain f

1 tt7soto/o202ss 26

l'

~

below that in the primary system until the

~

i$us,fortheoffsitedoseanalysis,itwasassumed s that.the s,L

)-

2. Conservative Assumptions f

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 Trip and Turbine Runback An earlier reactor trip is conservative for the offsite dose analysis, similar to the case for the overfill analy:is. Due to the assumed loss of offsite power, the condenser is not available

- for steam releases once tne reactor is tripped. Consequently, after reactor trip, steam is released to the atmosphere through the steam generator PORVs (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 i

modeling the Plant Vogtle reactor protection system, and this time j wasalsousedfortheoffsitedoseanalysis.(( l

-- a,e i

b. Steam Generator Secondary Mass If steam generator overfill do;es,ngt occur, a _,

results in a conservative i

117sr.10/020Ma 27

prediction of offsite doses. Thus, for the offsite dose analysis, the initial secondary mass was assumed to correspond to operation ,

- 0;c

~

c. AFW System Operation

-. Roc In Reference 2, it was determined that results in an increase in the calculated offsite radiation doses for aQGTR, whereas it was previously concluded that _

is conservative for the margin to overfill analysis.

However, it was also demonstrated in Reference 2 that ac dincethesinglefailureassumedfortheoffsiteradiation doseanalysgsis _

it is not necessary to assume an additional failure in the AFW system. Thus, the total AFW flow used for the margin to overfill analysis was also assumed for the offsite radiation dose ~

analysis. However, the delay time assumed for delivery of the AFW flow was , ,

d. Flashing Fraction When calculating the amount of break flow that flashes to steam,

- et c Since the tube  ;

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 - - c .

eYhus the assgmgtion that is conservative for the SGTR analysis.

1179<;1o/012:44 28

I j

3. Operator Action Times O

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,fortheoffsitedoseanalysisghe , _

at the time the ruptured steam generator is isolated. Before proceeding with the recovery operations, the

- a, c

~

Georgia Power Company has determined that an operator can[m .a, c -

Thus, it was assumed that the a, c - - u,c After the is

~

isolated, an additionsl delay time of 7 minutes (Table II.1) was assumed for the operator action time to initiate the RCS cooldown.

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. The transient results for this case are similar to the transient results for the overfill analysis until the time when the ruptured steam generator is isolated.

The transient behavior is different after this time-asine.e it is assumed that the' when theisolationisperfor5ed. ~

Following the tube rupture the RCS pressure decreases as shown in Figure III.1 due to the primary to secondary leakage. This depressurization results in reactor trip on an overtemperature delta-T

, signal. After reactor trip, core power rapidly decreases to decay heat levels and the RCS depressurization becomes more rapid. The 117seto/o12 ass 29

steam dug system is inoperable due to the assumed loss of offsite power, which results in the secondary pressure rising to the steam generator PORY setpoint as shown in Figure III.2. The decreasing pressurizer pressure leads to an automatic SI signal on' low pressurizer pressure. Pressurizer level also decreases more rapidly following reactor trip as shown in Figure III.3. After SI actuation, the SI flow rate exceeds the break flow rate, and the RCS pressure and pressurizer level increase until gn,c -

Major Operator Actions i 1. Identify and Isolate the Ruptured Steam Generator l

l The ruptured steam generator is assumed to be identified and isolated at 12 minutes after the initiation of the SGTR or when

! the narrow range level reaches 33%, whichever time is greater.

Since the time to reach 33% narrow range level is greater than 12 minutes, it was assumed that the ruptured steam generator is

~

l isolated at this time. The -

L ~

-slthistime.

a The failure causes the rupturedsteamgene7atortorapidlydepressurize,whichresultsin .

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 j decrease in the ruptured loop temperatures as shown in Figure III.4. The intact steam generator loop temperatures also decrease, as shown in Figure III.5, until the AFW flow is controlled to maintain the specified level. The decrease in the RCS temperatures results in an initial decrease in the RCS pressure and pressurizer level. However, the increased SI flow subsequently causes the RCS pressure and pressurizer level to increase again as shown in Figures III.1 and III.3, respectively. .

l l It is assumed that the time required for the operator to identify

a. , t.

that the - ,

t tmto/onone 30

i

\

el- i is 16 minutes. Thus, at 1704 l Iecondsthedepressurizationof"~therupturedsteamgeneratoris terminated.

2. Cool Down the RCS to establish Subcooling Margin

,4,4 After the a7 minute operator action time is imposed prior to initiation of  ;

cooldown. The depressurization of the ruptured steam generator  !

affects the RCS cooldown target temperature since the tegerature is dependent upon the pressure in the ruptured steam generator.

Since offsite power is lost, the RCS is cooled by dumping steam to the atmosphere using the intact steam generator PORVs. The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20'F plus an allowance of 24'F for instrument uncertainty. Because of the lower pressure in the ruptured steam generator, the associated temperature the RCS must be cooled to is also lower, which has the not effect of extending the time for cooldown. The cooldown is initiated at 2126 seconds

. and is completed at 3100 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 111.5.

The RCS pressure also decrsases during this cooldown process due to shrinkage of the reactor coolant as shown in Figure III.1.

3. Depressurize to Restore Inventory After the RCS cooldown, a 2 minute operator action time is included prior to depressurization. The RCS is depressurized at 3220 seconds to assure adequate coolant inventory prior to

!- terminating SI flow. With the RCPs stopped, normal pressurizwr spray is not available and thus the RCS is depressurized by opening a pressurizer PORV. The depressurization is continued until any of the following conditions are satisfied: RCS pressure 1178r.1D/0203:4 31

~

1 is less than the ruptured steam generator pressure and pressurizer level is greater than the allowance of 9% for pressurizer level -

uncertainty, or pressurizer level is greater than 69%, or RCS subcooling is less than the 24*F allowance for subcooling -

uncertainty. 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 111.3.

4. Terminate SI to Stop Primary to Secondary Leakage The previous actions have established adequate RCS subcooling, verified a secondary side heat sink, and restored the 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 24*F allowance for uncertainty, minimum AFW flow is available or at least one intact steam generator level is in the narrow range, the RCS pressure is ,

increasitag, and the pressurizer level is greater than the 9%

allowance for uncertainty. To assure that the RCS pressure is ,

i increasing, SI is not terminated until the RCS pressure increases j by at least 50 psi.

l After depressurization is completed, an cperator action time of 2 l minutes was assumed prior to SI termination. Since the above requirements are satisfied, SI termination is performed at this time. After SI termination, the RCS pressure decreases as shown l in Figure 111.1. The differential pressure between the RCS and the ruptured steam generator is shown in Figure III.S. Figure III.7 shows that the primary to secondary leakage continues after the SI flew is stopped until the RCS and ruptured steam generator ,

pressures equalize.

Iirstio/ciress 32 v- v- w-

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

For this case, the water volume in the ruptured steam generator is significantly less than the total steam generator. volume of 5906 ft 3 when the break flow is terminated. The mass of water in the ruptured steam generator is also shown as a function of time in Figure 111.9.

I G

l I

I t asrio/ot use 33 l

-m vs----- ----- -

l

~

TABLE III.1 PLANT V0GTLE SGT.1 ANALYSIS .

SE0VENCE OF EVENTS OFFSITERADIATIONDOSEANdLYSIS -

EVENT TIME (sec)

SG Tube Rupture O Reactor Trip 109 SI Actuated 308 Ruptured SG Isolated 740

_ _ 4.c 744 i

- -,a.c ,

1704 r RCS Cooldown Initiated 2126 l '

RCS Cooldown Terminated 3100 RCS Depressurization Initiated 3220

[

i i

RCS Depressurization Terminated 3310 SI Terminated 3430 [

! Break Flow Terminated 4638 1 i 1

i

. l l

117st10/012ssa 34

, _f a

V00TLE STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS RCS PRESSURE 2600.<

2422.

22:0. A E

{2000. (

w a

e 1:00.

S t -

e u

is::.

1420.

1200.

f S

i .,,,

i

"' c . teca. 2cce, seca. 4cas, seca.

! T!PC iSEC) i I t Figure !!!.1 RCS Pressure - Offsite Radiation Oose Analysis I

i-mortorosissa 35

i l V0GTLE STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYS!$

SECONDARY PRESSURE I4::.'

12 "' " '

INTACT SGs 5 ::::. s m

i S RUPTURED SG i

(

3

..,.,.. t, ,

d E

• 6::.

I

= -

i a I

• M :::.

i - ,

u, .

A-e.

. 1:::. 2:::. 5:::. st::. s:::.

Tirt ist:i f

Figure III.2 Secondary Pressure - Offsite Radiation Dose Analysis , [

1 I

l s

t 1179v 10/012844 36

V0GTLE STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS PRESSURIZER LEVEL 00.<

75.

70.<

65.-

a 62.

W D 55.

E U 50..

B y45.

s

. 40.<-

55, t .

I 1 50.

'O. 1020. 2203. 5020. 4203. 5000.

TIME ISCC) i Figure !!!.3 Pressurizer Level - Offsite Radis. tion Oose Analysis r

117sv.ie/otzess 37

V0GTLE STEAM GENERATOR TUBE RUPTURE

• OFFSITE DOSE ANALYSIS RUPTURED LOOP HOT AND COLO LEG RCS TEMPERATURES 650.

600. .

C THOT e

h$30.

k 5

a 602.

I $

I ge, 4

TCOLO v .

• 9 l

c'4: . -

8 '

bf l

l I

! '2 . 1000. 2000. 5000. 4220. 5000. .

l TIME ISECi Figure III.4 Ruptured Loop Hot and Cold Leg RCS Temperatures -

Offsite Radiation Dose Analysis p

6 11av.io/oitus 38

V0OTLE STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS INTACT LOOP HOT AND COLO LEO RCS TEMPERAiURES 650.

600.

- THOT a

g 550.

! jE00- TCOLD W

5

~

~ 450.

l j  ?

l h 4:2. -

e 552.

i 1

0. 1000. 2020. 5020. 42CC. SC20.

3 ,

1 fir'E ISEC1

]

6 Figure III.5 Intact Loon Hot and Cold Leg RCS Temperatures - .

Offsite Radiation Dose Analysis t I

1179v 1D/0128a4

. 39

J r

I V0OTLE STEAN GENERATOR TUBE RUPTURE F OFFSITE DOSE ANALYSIS  ;

O!FFERENTIAL PRESSURE BETWEEN RCS AND RUPTURED SO 2003.

1923.

1623.-

1400 -

l C 1200.-N U

l.

g :c:2.

I ....

,, ,, V .

, li ~

s s:a.

l

• l na.

220.-

I c.

'2" e . ie:a. me: . m a. .ees. s ::.

l itrt isc:i i

i I

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

Offsite Radiation Dose Analysis ,

i l

1179v:10/01:ssa 40

?

l

[

V0GTLE STEAN GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS PRIMARY TO SECONDARY BREAK FL0u 60.

50.

I.

40.- I C

Ms 3 50.

2

! 3

'a 20.

r z

0.

1 i

t a

2.

l I

i .

I '

i

. g g ,0. 1200. 2002. 5220. 4:20. 5000.

l l itet istti 9

~

Figure III.7 Primary to Secondary Break Flow Rate -

Offsite Radiation Dona Analysis h

1179v.10/012848 41

l

%7 i,

e V0GTLE STEAN GENERATOR TUBE RUPTURE  ;

i 0FFSITE 00SE ANALYSIS RUPTURED SG WATER YOLUME  ;

4253.'

• b 4:00.

T

_ 5750.

2 U 55 2.

I 8>

5252.

u ,

i 2!

I 3 :: . .

I

! 8;;, .

c w

j-I t

Ig 2E . .

5 G

! l 2250. i I

22:2.

t i

1750,0. 1020. 2222. 5220. 4220. 5:02. j j

r T!t'C 15CCI d

i i

Figure III.8 Ruptured SG Water Volume - Offsite Radiation -

't Dose Analysis i

t iI T

itntio/ciassa 42

__ _ __ _J

3 t

i

. l V0GTLE STEAM GENERATOR TUBE RUPTURE

<- OFFSITE DOSE ANALYSIS RUPTURED S0 uATER NASS f

22cate..

2:22:3.<

E d.19:000.

o f l

y l62::2.'

l I

N 4:020.4 I

~

g .

S u 122002. ,

I r

! 1C0200.'

I I I I

m "aa.

l *0 . 1000. 22:0. 5002. 4000. 5:02, 71"C ISCCI l

. Figure !!!.9 Ruptured SG Water Wass - Offsite Radiation Dose Analysis l

d i

i 1179v.lo/012:44 43 l

5. Mass Releases The mass releases were determined for use in evaluating the exclusion area boundary and low population zone radiation exposure. The steam releases from the ruptured and intact steam generators, the feedwater 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 hours 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 hours are used to calculate the radiation doses at the low population zone for the duration of the accident.

In the LOFTTR2 analyses, the SGTR recovery actions in Plant Vogtle Emergency Operating Procedure 19030-1 (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~

l should be cooled and depressurized to cold shutd g conditions.

~

i it was assumed that i the cooldown is performed using Plant Vogtle Emergency Operating Procedure 19033-1 (ERG ES-3.3), POST-SGTR C00LOOWN USING STEAM DUMP, since this method results in a conservative evaluation of the long j term mass releases for the offsite dose analysis.

The high level actions for the the post-SGTR cooldown method using steam dump in Plant Vogtle Emergency Operating Procedure 19033-1 (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.

1179t10/0128st 44

4 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 shutdm saisi~ n 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 Tesperature 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 since this minimizes the radiological releases and conserves feedwater supply. The PORVs 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 analysis, it was assumed that the c~ooldown is performed using steam dump to the

~

atmosphere via the intact steam generator PORVs. When the RHR I

system operating temperature is reached, the cooldown is stepped l

until RCS pressure can also be decreased. This ensures that pressure / temperature limits will not be exceeded.

k

3. Depressurize RCS to RHR System Pressure l

When the cooldown to RHR system temperature is completed, the pressure in the ruptured steam generator is decreased by releasing i

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 PORY en 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 depressurized by releasing steam via the PORV. As the ruptured steam generator 1179v;1o/012:ss 45 f

i

pressure is reduced, the RCS pressure is maintained equal to the pressure in the ruptured 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, a pressurizer PORY or auxiliary spray can be used to control RCS pressure if pressurizer spray is not available.

4. Cooldown to Cold Shutdown When RCS temperature and pressure have been reduced to the RHR system in-service values, RHR system cooling is initiated to complete the cooldown to cold shutdown. When cold shutdown conditions are achieved, the pressurizer can be cooled to terminate the event.

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

a. Nethodology for Calculation of Mass Releases .

The cperator acticns 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 leakage is terminated. l Following the termination of leakage, it was assumed that the RCS .

and intact steam ggerator conditions are maintained stable for a

~

until the cooldown is initiated. The PORVs for -

~

the intact steam generators were then assumed to be used to cool 1

l 1179v:1o/0128ts 46

down the RCS to the RHR system operating temperature of 350'F, at ,

the maximum allowable cooldown rate of 100'F/hr. The RCS and the intact steam generator tosperatures at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were then determined

. n.c The generator for the period from leakage terminatjon 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.c Since the ruptured steam generatorisisolated,nochangeintIIerupturedsteamgenerator condition *, 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 t'o be

. depressurized to the RHR in-service pressure of 390 psia via steam

- release from the ruptured steam generator PORY, since this maximizes the steam release from 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 flows 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 ,

ac

- Ihesteamreleasedfromtherupturedsteam

~

generatorfrom2to8hourswasdeterminedbasedon]

mov to/ciassa 47

i i

. Af ter 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, it is assuw) that further plant cooldown to cold shutdown as well as long-term cooling is provided by the RHR, *l system. Therefore, the steam releases to the atmosphere are i 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 ,

4 The mass release calculations were performed using the methodology discussed above. For the 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 condenser air djector. After l reactor trip, the releases to the atmosphere are assumed to be via j the steam generator PORVs. The mass release rates to the  ;

atmosphere from the LOFTTR2 analysis are presented in Figures l

!!!.10 and !!!.11 for the ruptured and intact steam generators, ,

i l

i respectively, for the time period until leakage termination. i

> 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 l to the atmosphere via the steam generator PORVs. The mass l

releases for the SGTR event for each of the time intervals j considered are presented in Table !!!.2. The mass releases prior to break flow termination, from break flow termination until l l

l 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 l l results indicate that approximately 101,900 lbe of steam are released from the ruptured steam generator to the 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 169,000 lbe of primary water is I f

transferred to the secondary side of the ruptured steam generator .

f before the break flow is terminated.  !

i l

[

i msaofoissas 48

. TA8LE 111.2 PLANT V0GTLE SGTR ANALYSIS

. NASS RELEASES OFFSITE RADIATION DOSE ANALYSIS TOTALMASSFLOW(POUN05)

TINE PERIOD 0-TRIP TRIP - TTBRK - T2 HRS -

TTBRK T2 HRS TRHR Ruptured SG Condenser 118,100 0 0 0 Atmosphere 0 101,900 0 33,900 Feedwater 113,400 51,400 0 0 Intact SGs Condenser 351,200 0 0 0 Atmosphere 0 292,400 220,300 895,400 Feedwater 351,200 529,100 235,000 900,300 Break ' Flow 4,800 164,200 0 0 l

TRIP = Time of reactor trip = 109 sec.

TTBRK = Time when break flow is terminated = 4638 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 RiiR 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.

• l

~

l i

1179v;1o/0201:a 49

i TABLE !!!.3 PLANT V0GTLE SGTR ANALY$!$ ,

$IANARIZED MASS RELEASES OFFSITE RADIATION DOSE ANALYS!$

TOTALMASSFLOW(P00N05) -

i 0- TT8RK - 2 HRS -

TTBRK- 2 HRS 8HR$

Ruptured SG Condenser 118,100 0 0 ,

Atmosphere 101,900 0 33,900 -

'l Feedwater 164,800 0 0  :

1 Intact SGs ', ,

Condenser 351,200 0 0 .

l Atmosphere 292,400 220,300 895,400

?

j Feedwater 880,300 235,000 900,300 j Break Flow 169,000 0 0 I

2  !

i i

I h

i t

1179v:10/01:ss: 50 ,

. V0GTLE STEAM OENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS RUPT SG ATMOSPHERIC MASS RELEASES 700.<

y620.' -

s 5

g500.' ,

U '

W o 422.-

Y M

g500.

I 223.-

8 h102.

(

8. 1928. 2880. 5220. 4482. 5C30.

TIMC IE C)

Figure !!!.10 Ruptured SG Wass Release Rate to the Atmosphere -

Offsite Radiation Dose Analysis 11 m io/oastas 51

T V0GTLE STEAM GENERATOR TUBE RUPTURE UFFSITE DOSE ANALYSIS INTACT SGS ATMOSPHERIC MASS RELEASE (LB/SEC) 2e99.'

G tese.

M s

y 1see.

M c 14ec.

2

• W

• g 1222.

E e icac..

=

W 9ee. .

i

" 6ee.

I w '

- 4ee. '

h' E

- 2ec..  ;

Q-i

' 's . tees. 2ees. seee. 4eee, sees.

T!!9C ISCCl I

l-Figure !!! 11 Intact SG hans Release Rate to the Atmosphere - -  !

Offsite Radiation Dose Analysis i E

timio/nasies 52

i

8. Offsite Radiation Dose Analysis The evaluation of the radiological consequences of a steam generator tube rupture event assumes that the reactor has been operating at the maximum allowable Technical Specification limit for primary coolant activity and

primary to secondary leakage for sufficient time to establish equilibrium concentrations of radionuelldes in the reactor coolant and in the

secondary coolant. Radionuclides from the primary coolant enter the steam l 1 generator,viatherupturedttie,andarereleasedtotheatmosphere through the steam generator PORVs and safety valves and via the condenser  !

airejectorexhaust.

The quantity of radioactivity released to the environment, due to a SGTR, .

I 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 i turbine condenser hot well. All of these parameters were conservatively evaluated for a design basis double ended rupture of a single tube.

1. Design Basis Analytical Assumptions ,

The major assumptions and parameters used in the analysis are itemized  ;

in Table !!!.4.

i i 2. Source Term Calculations The radionuclide concentrations in the 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 preaccident and accident initiated iodine spikes. l e

t j 117klo/canas 53

i. Accident Initiated Spike - The initial primary coolant iodine concentration is 1 uCi/gs 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 corresponding to the initial primary system iodine _

concentration. The duration of the spike, .

is sufficienttoincreasetheinitialRCSI-13Iinvento7ybya factor of . -

11. Preaccident Spike - A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine concentration from 1 to 60 uCi/ gram of D.E. I-131.

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

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

l 3. Dese Calculations .

The iodine transport model utilized in this analysis was proposed by Postma and Tan (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 atomi:ation.  !

Removal credit is taken for scrubbing of iodine contained in the <

atomized coolant droplets when the rupture site is below the secondary i

I water level. The fraction of primary coolant iodine which is not assumed to become airborne imediately mixes with the secondary water and is assumed to become airborne at a rate proportional to the

~

steaming rate and the iodine partition coefficient. This analysis

. l m er.io o m aa 54 l

l

\

conservatively assumes an iodine partition coeffici9nt of 100 between the steam generator liquid and steam phases. The model takes no scrubbing or mixing credit when the rupture site is above the secondary water level. Droplet removal by the dryers is conservatively assumed to be negligible. The iodine transport model is illustrated in Figure !!!.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 !!!.2.

b. The time dependent fraction of rupture flow that flashes to steam and is immediately released to the environment is presented in Figure !!!.13.

. c. In the iodine transport model, the time dependent iodine removal efficiency for scrubbing of steam bubbles as they rise from the rupture site to the water surface assumes that the rupture is located at the interser, tion of the outer tube row and the upper anti-vibration bar. However, in accordance with the methodology in Reference 2, the tube rupture break flow was conservatively calculated assuming that the break is at the top of the tube sheet. The water level above the top of the tubes in the ruptured and intact steam generators is shown in Figure !!!.14. The iodine removal efficiency is determined by the method suggested by Postma and Tam (Ref. 4). The iodine removal efficiencies are shown in

. Figure !!!.15.

1179v.lo/0201:a 55

d. During the time period that the rupture site is uncovered, all of ,

]

the activity carried by the break flow is assumed to be directly released to the environment, i.e., the activity carried by the break flow will neither six with the secondary water nor "

partition. The rupture site is considered to be covered when the i secondary water level is approximately 12 inches over the rupture location (approximately 10 inches over the apex of the tube bundle).

e. The total primary to secondary leak rate is assumed to be 1.0 gpa as allowed by the Technical Specifications. The leak rate is assumed to be 0.70 gpa to the three intact steam generators and 0.30 gpm to the ruptured steam generator.

f. The iodine partition factor between the liquid and steam of the ruptured and intact steam generators is assumed to be 100,

g. 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 bound:ry of the low .

l population zone.

h. Short-term atmospheric dispersion factors (x/Os) for accident analysis and breathing rates are provided in Table !!!.7. The breathing rates were obtained from NRC Regulatory Guide 1.4,

! (Ref. 5).

i

4. Offsite Thyroid Dese Calculation Model Offsite thyroid doses are calculated using the equation:

( \-

D

• 0U g IAR)gj Th .

(BR)3 (x/Q)3)_ ,

6 l

i 1179v:10/02048: 56

I where ,

i l

integrated activity of iodine nuclide i released

, (IAR)$3 =

during the time interval j in Ci*

(BR)j

= breat ing rate during time interval .] in q meter /second (Table !!!.7) .

= atmospheric dispersion factor during time (x/0)) ]

interval j in second/ meter 3(TableIII.7) l (DCF)q

= thyroid dose conversion factor via inhalation for iodine nuclide i in. rem /Ci (Table III.8) l L D Th

= thyroid dose via inhalation in rem l

S. Results l

. 1 Thyroid doses at the Exclusion Area Boundary and Low Population Zone l are presented in Table 111.9. All doses are well within the allowable guidelines as specified by Standard Review Plan 15.6.3 and 10CFR100.

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 thi low population zone.

117sv.tomaiss 57

i TABLE III.4 PLANT V0GTLE SGTR ANALYS!$

PARANETERS USED IN EVALUATING RADIOLOGICAL CONSEQUENCES I. Source Data .

A. Core power level, NWt 3565 B. Total steam generator tube 1.0 leakage, prior to accident, gpa C. Reactor coolant iodine j activity:

1. Accident Initiated Spike The initial,RC iodine activities based on 1  !

, uCi/ gram of 0.E. 1-131 .

are presented in

, Table III.S. The iodine -

, appenrance rates assumed for the accident initiated spike are presented in Table III.6.

i

2. Pre-Accident Spike Primary coolant iodine activities based on 60 ,

uCi/ gram of D.E. 1-131 i are presented in [

Table III.S. t D. Secondary system initial activity Dose equivalent of 0.1  !

uCi/gm of I-131, I presented in Table !!!.5.  !

l l

\ ,

tuotto/ozoin 58

TABLE!!!.4(Sheet 2)

E. Reactor coolant mass, grams 2.3 x 10 8 F. Initial Steam generator mass 4.0 x 10 7  !

(each), grams G. Offsite power Lost at time of reactor trip H. Primary-to-secondary leakage 8 duration for intact SG, hrs.

t I. Species of iodine 100 percent elemental i

!!. Activity Release Data A. Ruptured steam generator t

1. Rupture flow See Table !!!.2

2. Rupture flow flashing fraction See Figure !!!.13  !

3. Iodine scrubbing efficiency See Figure !!!.15  ;

4. Total steam release, lbs See Table !!!.2 t

5. Iodine partition factor 100 i Location of tube rupture Intersection of outer

^

6. L tube row and upper l anti-vibration bar t

I

! un munu 59

i 2 i TABLE!!!.4(Sheet 3)

8. Intact steam generators

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

2. Total steam release, 1bs See Table !!!.2 P

3. todine partition factor 100 -

C. Condenser l l

1. Iodine partition factor 100 i i

D. Atmospheric Dispersion Factors See Table !!!.7  :

S u

I .

l 4

i l

J P

tir w etozoisa 60

. TABLE !!!.5 PLANT V0GTLE SGTR ANALYSIS .

4 1

IODINE SPECIFIC ACTIVITIES IN THE PRINARY AND SECONDARY COOLANT BASED ON 1. 60 AND 0.1 uCi/oras 0F 0.E. I-131 Specific Activity (uCi/on)

Primary Coolant Secondary Coolant Nuclide 1 uti/ou 60 uCi/ou 0.1 uCi/ou I-131 0.76 45.6 0.076 '

i 1-132 0.76 45.6 0.076 i 1-133 1.14 68.4 0.011

! I-134 0.'20 11.7 0.020 l

l-135 0.63 37.8 0.063 l  !

i l

f i

I itnv.ie/cacin 61 i

i

. TABLE !!!.6 PLANT V0GTLE $6TR ANALYSIS 100!NE SP!KE APPEARANCE RATES -

(CURIES /SECOND)

!-131 1-132 1-133 I-134 I-135 1.7 9.0 3.6 5.5 3.4 i

t i

f i

}

2 t

l i

i

, i

[

I l

l l

a 4 I i

i

\ I

{

11 W 10/0201M 62

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

4 TABLE III.7

. PLANT V0GTLE SGTR ANALYSIS ATHOSPHERIC DISPERSION FACTORS AND BREATHING RATES Time Exclusion Area Boundary Low Population Breathing 3 3 3 (hours) x/Q(Sec/m) Zone x/Q (Sec/m ) Rate (m/Sec)(4]

-4 0-2 1.8 x 10'4 7.2 x 10 -5 3.47 x 10 2-8 -

3.3 x 10 -5 3.47 x 10 -4 1

I 1179v:10/020144 63

TABLE III.8 PLANT V0GTLE SGTR ANALYSIS -

THYROID DOSE CONVERSION FACTORS (Rem / Curie)(Ref.6)

Nuclide I-131 1.49 x 10 6 I-132 1,43 x 10 4 I-133 2.69 x 10 5 I-134 3.73 x 10 3 I-135 5.60 x 10 4 I

l 1

l l

e 9

1179v:1D/020184 64 l .-- . ._. . . . _ - . _

TABLE III.9 PLANT V0GTLE SGTR ANALYSIS 0FFSITE RADIATION DOSES Thyroid Doses (Ree)

Calculated Allowable Value Guideline Value

1. Accident Initiated Iodine Spike ExclusionAreaBoundary(0-2hr.)

Thyroid 4.8 30 Low Population Zone (0-8 hr.)

Thyroid 1.9 30

2. Pre-Accident lodine Spike

. Exclusion Area Boundary (0-2 hr.)

Thyroid 31.4 300 LowPopulationZone(0-8hr.)

Thyroid 12.6 300 e

• O O

tuer.to/ozom 65 i

D DROPLETS NOT SCRuesto FLASH -

INTO M SCRUBBING '

VApon 6 PMMARY g3 DN S

BREAK -

scmueeto T COVERED 7 u DRoes g WATER A NOT

• N A Ft. ASHED , SECONDARY ,

- 0 T WATEM 31y m 5 y P O

/No A $

p C '

C H

• = E FLASH R INTO -

' g

• varon k DROPLET 5 SPRAY NOT g SRTAKU P FU.CNED INTO OROPS ,

'] _

ravinoN Figure 111.12 Iodine Transport Model - Offsite - '

Radiation Dose Analysis 1179v:10/020184 g

7 e

V0OTLE STEAN GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS BREAK FLOW FLASHING FRACTION

.16<

/

.14' f.10' 8

E .1-

.00 r i

k C .i6-M W \

m .e4- j,

.22

'O . 1020. 2000. 5200. 4200. 5000.

T!!T ISEC)

Figure III.13 Break Flow Flashing Fraction - Offsite Radiation Dese Analysis 1179v;1D/020144 67

^

V0GTLE STEAM GENERATOR TUBE RUPTURE OFFSITE OOSE ANALYSIS SG WATER LEVEL ABOVE TOP OF TUBES 240.

220.

k200.

RUPTURED SG

$ 180.

B

~

160.

b '

, g140. INTACT SGs y120. .

[ '

! 2

100. '

l l W I

d ea.

! 5 g 60.

$ 40.

f 20.

• 0. 1000. 2000. 5000. 4000. 5000.

TIME ISECl l

l l

Figure 111.14 SG Water Level Above Top of Tubes -

Offsite Radiation Oose Analysis 1

! mkiotozoiss 68

l I

i e

0.03 t

V0GTLE SGTR W

W 0.02 S

w

!0.01 -

6 m

, I '

i i , i g, f -- .

3 0 500 1000 1500 2000 2500 .

TIME (SECONDS}

, Figure III.15 Iodine Scrubbing Efficiency - Offsite Radiation Dose Analysis .

11 n v:1otozoiss 69

IV. CONCLUSION An evaluation has been performed for a design basis SGTR event for the Plant Vogtle Units 1 and 2 to demonstrate that the potential consequences are acceptable. An analysis was performed to demonstrate margin to steam generator overfill assuming the limiting single failure relative to overfill.

The limiting single failure is the failure of a.C The

~

results of this analysis indicate that the recovery actions can be comp 1sted 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 th limiting eingle failure for offsite coses. For this analysis, it was assumsd enat the a,c Tho ,

~

primary to secondary break flew and the mass releases to the atmospheru cre determined for this case, and the offsite radiation doses were calculated

• using this information, The resulting doses at the exclusion area boundary l and low population zone are well within the allovable guidelines as specified l by Standard Review Plan 15.6.3 and 10CFR100. Thus it is concluded that tna consequences of a design bssis steam generater tube rupture at Plant Vogtle would be acceptable.

1 l

l l

1179v:1o/0201ss 70 l

1

, V. REFERENCES

• 1. Lewis, Huang, Behnke, Fittante, Gelman, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," WCAP-10750-A, August 1987.

l

2. Lewis, Huang, Rubin, "Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident," Supplement 1 to WCAP-10750-A, March 1986.

3. Southern Company Services, "SGTR Event Operator Action Times Using Vogtle Simulator", NCA-7002, February 1987.

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

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

c

6. NRC Regulatory Guide 1.109, Rev. 1, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix 1", October 1977.

Y 1-e 117 w o/o201ss 71

, r . .,

COMPREHENSIVE EQUIPMENT LIST Fcr Procedure 19030-1 Rev. 4 10-26-87 EOP Ref. Step Principal Equipment Safety Grade Backup Equipment / Guideline Safety Grade

1. Check if RCPs should RCP Circuit Br(akers o assume LOSP be stopped.

RCP No. Breaker j

1-1201-P6-001-M01 252-1NAA08 N 252-1AAA Y 1-1201-P6-002-M01 252-INAB06 N 252-1BAB Y @

1-1201-P6-003-M01 252-lNAA07 N .p 252-1CAC Y @

1-1201-P6-004-M01 252-INAB07 N $

252-1 DAD Y w

I 2. Identify ruptured o SG narrow range level Y SGa by following o High radiation detectors Y

, conditions. for SC steamlines (RE-13119, l 13120, 13121 and 13122) l l 3. Icciate flow from o SC ARV (PV-3000, 3010, 3020, Y Remote Manual switch in control room Y motive and indication power supply Y ruptured SGs. 3030)

Motive and indication power Y Manually cose the SG ARV Block Valves Y l supply (001, 002, 003, 004, 136, 137, 138 and 139) l l

o Steam supply valves from o Manual Valves Y ruptured SG to TDAFW pump 1301-005 l

! 1301-HV-3009/125 Vdc MCCIBD1M Y/Y 1301-007 1301-HV-3019/125 Vdc MCCIAD1M Y/Y o Verify blowdown isolation o Manually shut valve HV-7603A, B, C, Y valve from ruptured SG is or D with handwheel '

shut (HV-7603A, B, C, or D) Y

-e 125 Vdc 1AD12 or IBD12 Y 4 J

l l

1

. ]l

EOP Ref. Step Principal Equipment Safety Grade Backup Equipment / Guideline Safety Crade o Shut MSIV and bypass valvec o Steam Dump Valves on ruptured SG 1TV 0500A/1AD11 & IBD12 Y/Y Vogtle is unique for a PWR B/1AD11 & IBD12 Y/Y because it has two MSIVs in C/1AD11 & IBD12 Y/Y series on each steam line. D/JAD11 & IBD12 Y/Y 1HV-3006A/125Vdc 1AD12 Y/Y E/1AD11 & 1BD12 Y/Y B/125Vdc 1BD12 Y/Y F/1AD11 & 1BD12 Y/Y 1HV-3016A/125Vdc 1AD12 Y/Y C/1AD11 & IBD12 Y/Y B/125Vdc IBD12 Y/Y H/1ADil & IBD12 Y/Y llIV-3026A/125Vdc 1AD12 Y/Y J/1AD11 & JBD12 Y/Y B/125Vdc IBD12 Y/Y 1PV 0507A/1AD11 & IBD12 Y/Y IHV-3036A/125Vdc 1AD12 Y/Y B/1AD11 & IBD12 Y/Y B/125Vdc IBD12 Y/Y C/]AD11 & IBD11 Y/Y

~

Each steam line has two bypass valves in series.

1HY-13005A/125Vdc 1AD12 Y/Y IHY-13006A/125Vdc 1AD12 Y/Y IHY-13007A/125Vdc 1AD12 Y/Y 1HY-13008A/125Vdc L@l2 Y/Y IHY-13005B/125Vdc 1BD12 Y/Y IHY-13006B/125Vdc 1BD12 Y/Y IHY-13007B/125Vdc 1BD12 Y/Y IHY-13008B/125Vdc 1BD12 Y/Y o Condenser sparger valves HV-6194A/480V MCC INBB N/N HV-6194B/480V MCC INBB N/N o Steam jet air ejector valves HV-4084B/480V MCC INBP N/N HV-4C85B/480V MCC 1NBP N/N 1

o MSR steam supply control switches HS-6030 (shuts HV-6030 and IIV-6179) N 480V MCC 1NBL N i'

HS-6015 (shuts HV-6015 and IIV-6181) N 480V MCC INBN N 4 . _ . - - __

EOP Ref. Step Principal Equipment Safety Grade Backup Equipment /Cuideline Safety Grade o Shut all remaining MSIVs and bypasses (power sources previously given step 3) Y/Y o Intact SC ARVs for steam dump (power sources given step 3) Y/Y o (Procedure 19131-1)

ECA - 3.1 Guideline

4. Check ruptured SCs o SC narrow range level (see Y Icvels step 2) o Stop feed flow to ruptured SG.

AFW flow control valves: o AFW Pumps HV-5120, 5122, 5125, 5127 Y 1-1302-P4-002-M01 (4160V bus IBA03) Y/Y 125Vdc MCC 1CDIM Y 1-1302-P4-003-M01 (4160V bus 1AA02) Y/Y HV-5132, 5134/480V MCC IBBB Y/Y HV-5137, 5139/480V MCC 1ABB Y/Y

5. Check PRZR PORVs o PRZR PORV o Parallel PORVs in series with block Y cnd block valves 455A/125Vdc MCC 1AD1M Y/Y valves 456A/125Vdc MCC IEDIM Y/Y o (Procedure 19131-1)

ECA-3.1 Guideline o PRZR PORV block valves

. 1HV-8000A/480V MCC LABE Y/Y 1HV-8000B/480V HCC IBBE Y/Y

6. Check SCs secondary o Steam generator pressure o (Procedure 19020-1) pressure boundaries transmitters E-2 Cuideline PI-514, 515, 516 Y Pr-524, 525, 526 Y l PI-534, 535, 536 Y PI-544, 545, 546 Y l

w EOP Ref. Step ' Principal Equipment Safety Grade Backup Equipment / Guideline Safety Grade' i 7. Check intact SG 1evels o SG narrow range level Y (see step 2) o AFW control valves Y o AFW pumps (see step 4) Y.

(see step 4)

8. R

set SI o SI reset device Y

9. RIset containment o Check containment area Isolation Phase A radiation monitors 2

RE-0005 (scheme 11CQRM2) Y RE-0006 (scheme 12CQRM2) Y l

o Containment inclation Phase A .Y j reset device i 10. Establish instrument o Instrument air system N o Service air system N air to containment

o HV-9378/125Vdc 1AD12 Y/Y

11. Varify all AC buses o Controls to restore offsite Y o Emergency diesel generators Y i (energized by offsite power j power)

12. Check if RHR pumps o RCS pressure (PT-403, 405, Y j chould be stopped 408, 418, 428, 438)

I o RHR pump switches Y l

t 13. Check ruptured SG o SG pressure transmitters Y. o SG ARVs (see step 3) LY pressure (see step 6) (Procedure 19131-1) i

- ECA-3.1 Guideline

! 14. Initiate RCS cooldown o Steam dump valves Y j (see step 3) l o Core exit thermocouples Y o (Procedure 19131-1) l ECA-3.1 Guideline i

l  !

~.

EOP Ref. Step Principal Equipment Safety Grede Backup Equipment /Cuideline Safety Grade

15. Check ruptured SCs o SC pressure transmitters Y o (Procedure 19131-1) prssoure (see step 6) ECA-3.1 Guideline

16. Check RCS Subcooling o RCS subcooling monitor Y o (Procedure 19131-1)

Monitor Indication ECA-3.1 Guideline

17. Depressurize RCS to o PRZR spray valves.PV-04553,C N o Stop RCPs SeeLStep 1-minimize break flow cad refill PRZR o RCS pressure Y (see step 12) o SG pressure Y (see step 6) o PRZR level Y Ur-459, 460, 461 o RCS subcooling monitor Y (see step 16)

18. Depressurize RCS o PRZR PORV Y o Auxiliary spray valve for PRZR N using PRZR PORV (see step 5) HV-8145 o RCS pressure Y (see step 12) o PRZR level Y o (Procedure 19133-1) ,

(see step 17) ECA-3.3 Guideline) o SG pressure Y (see step 6) o RCS subcooling monitor Y (see step 16)

19. Check RCS pressure o RCS pressure monitoring Y o PORV block valve Y (see step 12) (see step 5)

- e EOP Ref. Step Principal Equipment' Safety Grade Backup Equipment /Cuideline Safety Grade' o PORV valve status indication (same Y source as control power - see step 5) o PORV discharge line temperature N o (Procedure 19101-1)

ECA-3.1 Guideline

20. Check if ECCS flow o RCS subcooling monitor Y o (Procedure 19131-1) chould be terminated (see step 16) ECA-3.1 Guideline o AFW flow transmitters

' train l' - Fr-5150, 5152, Y 15151, 15153 "train 2" - FI-15150, 15152 Y 5151, 5153 o SC narrow range level Y (see step 2) o RCS pressure (see step 12) Y o PRZR level (see step 17) Y

21. Stop ECCS pumps and o SI pump switches Y place in standby

22. Establish charging o HV-8111A/480V MCC IBBD Y/Y

. flow HV-8111B/480V MCC 1BBB Y/Y

a. open CCP normal o HV-8110/480V MCC LABD Y/Y miniflow

b. shut CCP alternate o HV-8508A/480V MCC 1ABB Y/Y miniflow HV-85083/480V MCC IBBB Y/Y

c. shut seal flow o HV-182 N

~ '7 EOP Ref. Step Principal Equipment Safety Grade Backup' Equipment / Guideline Safety Grade

d. open charging o HV-8105/480V MCC IBBB '

Y/Y line HV-8106/480V MCC 1ABR Y/Y

e. shut Brr discharge o HV-8801A/480V MCC LABD Y/Y- -

isolation valves HV-8801B/480V MCC IBBD Y/Y ,

'~

(Verhal description -

of valves will ,  ;

differ for Unit 2) ,-

o CVCS charging pumps -

Y o Safety injection pumps Y NorE The principal equipment used in the remaining steps is not described since at this point in the accident the RCS and SC pressure should be equal. .

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