Text

Nonproprietary LOFTTR2 Analysis for Steam Generator Tube Rupture for South Texas Project Units 1 & 2
	ML19325D122
Person / Time
Site:	South Texas
Issue date:	09/30/1989
From:	Holderbaum D; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML19325D119	List: ML19325D120; ML19325D122; ST-HL-AE-3236, Forwards Nonproprietary WCAP-12370 & Proprietary WCAP-12369, LOFTTR2 Analysis for Steam Generator Tube Rupture for South Texas Project Units 1 & 2. Analysis Shows That Calculated Offsite Doses within NUREG-0800,SRP & 10CFR100 Limits;
References
	WCAP-12370, NUDOCS 8910190026
	Download: ML19325D122 (82)
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WESTINGHOUSE CLASS 3 WCAP-12370 LOFTTit ANALYSIS FOR A STEAN SENERATOR TUBE RUPTURE FOR THE SOUTH TEXA3 PRO X CT UNITS 1 AND 2 L

D. F. Holderbaum

t. N. Lewis K. Rubin l

1 l

SEPTEW ER 1989 Nuclear Safety Department l-l

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Westinshouse Electric Corporation fluelear Energy Systems P.O. Box 355 Pittsb h, Pennsylvania 15D0 I') 1988 by No inghouse Electric Corporation sowie,wome

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1 k TABLE OF CONTENTS l

I f121 0

1.

INTRODUCTION 1

ANALYSIS OF MARGIN TO STEAM GENERATOR OVERFILL 4

A.

Design Basis Accident 4

B.

Conservative Assumptions 5

C.

Operator Action Times 7

D.

Transient Description 14

)

!!!. ANALYSIS OF,DFFSITE RADIOLOGICAL CONSEQUENCES 26 A.

Thermal and Hydraulic Analysis 26 f

l'.

Design Basis Accident 26 2.

Conservative Assumptions 27 i

3.

Operator Action Times 29 l

4.

Transient Description 30 l

5.

Mass Releases 45 t

B.

Offsite Radiation Dose Analysis 54 IV.

CONCLUSION 75 V.

REFERENCES 76

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LIST OF TABLES i

l I&ltla 11114 Fast i

II.1 Operator Action Times for Design Basis Analysis 13

!!.2 Sequence of Events - Margin to Overfill Analysts 19 l

t

!!!.1 Sequence of Events - Offstte Radiation Dose Analysts 35

!!!.2 Mass Releases - Offsite Radiation Dose Analysis 50 III.3 Susuarized Mass Releases - Of'fsite Radiation Dose

$1 Analysis l

!!!.4 parameters Used in Evaluating Radiological 62 Consequences

!!!.5 lodine Specific Activities in the Primary and secondary 65 l

l Coolant l

!!!.6 lodine Spike Appearance Rates 66 l

III.7 Noble Gas Specific Activities in the Reactor Coolant 67 Based on 11 Fuel Defects I

l i

!!!.8 Atmospheric Dispersion Factors and Breathing Rates 68 III.9 Thyroid Dose Conversion Factors 69 i

III.10 Average Gasma and Beta Energy for Noble Gases and Iodines 70

)

III.11 Offsite Radiation Doses 71 l

2036v:10/030789 11

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LIST OF FIGURES Fieure' Title P,gge i

'!!.1 Pressuriser Level - Margin to Overfill Analysis 2D

!!.2 RCS Pressure - Margin to Overfill Analysis 21

)

!!.3 Secondary Pressure - Margin to Overfill Analysis 22

!!.4 Intact Loop Hot and Cold Leg RCS Tenperatures -

23 Margin to Overfill Analysis 11.5 Primary to Secondary Break Flow Rate - Nargin to 24

)

Overfill Analysis J

11.6 Ruptured SG Water Volume - Margin to Overfill Analysis 25 1

I 111.1 RCS Pressure - Offsite Radiation Dose Analysis 36 l

III.2 Secondary Pressure - Offsite Radiation Dose Analysis 37 j

111.3 Pressuriser Level - Offsite Radiation Dose Analysis 38 1

l 111.4 Ruptured Loop Hot and Cold Leg RCS Temperatures -

39 Offsite Radiation Dose Analysis

)

111.5 Intact Loop Hot and Cold Leg RCS Tenperatures -

4D Offsite Radiation Dose Analysis

!!!.6 Differential Pressure Between RCS and Ruptured 41 SG - Offsite Radiation Dose Analysis l

111.7 Primary to Secondary Break Flow Rate - Offsite 42 Radiation Dose Analysis i

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LIST OF FIGURES (Cont) 1 1

Pa2' Figure Title a

1

!!!.8 Ruptured SG Water Volume - Offsite Radiation Dose 43 Analysis

!!!.9 Ruptured SG Water Mass - Offsite Radiation Dose Analysis 44

' !!!.10 Ruptured SG Nass Release Rate to the Arsosphere -

52 j

Offsite Radiation Dose Analysis

]

1 111.11 Intact SGs Wass Release Rate to the Atmosphere -

53 Offsite Radiation Dose Analysis l

111.12 lodine Transport Model - Offsite Radiation Dose Analysis 72 1

111.13 Break Flow Flashing Fraction - Offsite Radiation 73 Dose Analysis l

111.14 SG Water Level Above Top of Tubes - Offsite 74 Radiation Dose Analysis

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i 2036v:10/0907a0 iy

I.

INTRODUCTION An evaluatien for a design basis steam generator tube rupture (SGTR) event has been performed for the South Texas Project (STP), 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 doses are less than the allowable guidelines.

The South Texas Project employs two essentially identical Westinghouse pressurized water reactor (PWR) units rated at 3800 MWt. The reactor coolant system for each unit has four reactor coolant loops with Model E2 steam generators. Since the reactors, structures, and all auxiliary equipment are substantially identical for the two units, the SGTR evaluation is applicable for both units.

It is also noted that both units are currently licensed to 1

operate with Westinghouse standard fuel with a negative moderator temperature coefficient. However, it is anticipated that the technical specifications will be changed to permit operation with a positive moderator temperature cwfficient for future fuel cycles. Therefore, a more limiting positive moderator temperature coefficient was assumed for the SGTR evaluation such that the results are applicable for the current licensing basis as well as for future operation with a positive moderator temperature coefficient. The evaluation is also applicable for up to 15 percent steam generator tube plugging and for a minimum auxiliary feedwater flow rate of 500 ppm per steam generator.

The steam ponerator tube rupture analyses were performed for South Texas using the methodology developed in WCAP-10698 (Reference 1) and Supplement I to WCAP-10698 (Reference 2). This 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 March 30, 1987. The LOFTTR2 program, an updated version of the LOFTTR1 program, was used to l

perform the SGTR analysis for South Texas.

The LOFTTR1 program was developed as part of the revised SGTR analysis methodology and was used for the SGTR l

evaluations in References 1 and 2.

However, the LOFTTR1 program was l

subsequently m'odified to accommodate ste:m generator overfill and the revised 2036v:10/o007a0 1

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program, designated as LOFTTR2, was used for the evaluation of the consequencesofoverfillinWCAP-11002(Reference 3). The LOFTTR2 program is identical to the LOFTTR1 program, with the exception that the LOFTTR2 program j

' has the additional capability to represent the transition from two regions (steam and water) on the secondary side to a single water region if overfill j

occurs, and the transition back to two regions again depending upon the l

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.

Plant response to the SGTR 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 analysis methodology includes the simulation of the operator actions for recovery from a steam generator tube rupture based on the South l

Texas Emergency Operating Procedures (EOPs), which were developed from the Westinghous'e Owners Group Emergency Response Guidelines (ERGS).

In subsequent references to the South Texas E0Ps, the specific 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 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 generator overfill since this could potentially result in a significant 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 analysis was also performed to determine the offsite radiation doses, assuming the limiting single failure relative to offsite doses without steam generator overfill. The limiting single failure assumptions for these analyses are consistent with the methodology in References 1 and 2.

203sv:1o/0e07:9 2

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For the margin to overfill analysis, it was assumed that the [

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LOFTTR2 analysis to determine the margin to overfill was performed for the time period from the tube rupture until the primary and secondary pressures era 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. The results of this analysis demonstrate that there is margin to steam generator overfill for South Texas.

Since steam generator overfill does not occur, the results of the offsite l

l radiation dose analysis represent the limiting consequences for South Texas.

l For the analysis of the offsite radiation doses,

~6toThe primary to secondary break flow and the steau releasestotheatmosphrefromboththerupturedandintactsteamgenerators l

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 f

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 maximum allowable Technical' Specification limit prior to the accident. The results of,

this analysis show that the offsite doses for South Texas are within the allowable guidelines specified in the Standard Review Plan, NUREG-0800, Section 15.6.3, and 10CFR100.

l 203sv;1o/De07a0 3

i II. ANALYSIS OF MARGIN TO STEAN GENERATOR OVERFILL i

An analysis was performed to determine the, margin to steam generator overfill 4

for a design basis SGTR event for South Texas. The analysis was performed using the LDFTTR2 program and the methodokgy 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 l

The accident modeled is a double-ended break of one steam generator tube h

locatedatthetogIhelocationofthebreak of the, tube sheet a, e It 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.

The most limiting single failure with respect to steam generator overfill

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was determined to be

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single failure for the four-loop South Texas plants is a, e TheSouthTexasAFWsystemconsistsoffourindependenttrains(three identical motor-driven pumps and one turbine-driven pump of equal capacity) with each feeding a dedicated steam generator. There is an AFW flow control valve for each steam generator in the flow path from the associated AFW pump. The AFW flow control valves would be normally open and are used to terminate feedwater flow to the ruptured steam generator,e and control inventory in the intact steam generators.

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.D.

[mervativeAssumptions

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

reported in Reference l'.

The conservative conditions and assumptions L

which were used in Reference 1 were also used in the LOFTTR2 analysis to determine the margin to steam generator overfill-for South Texas with the 1

L exceptit:, of the following differences.

L l

1.

Reactor Trio and Turbine Runback D

A turbine runback can either be initiated automatically or the.

operator can manually reduce the turbine load following an SGTR to t

attempt to prevent.a reactor trip. For the reference plant analysis

. in WCAP-1069A,s. reactor trip was calculated to occur at approximatel -li,e E,d turbine runback to an was simulatedInsedonarunbackrateof" fhaeffectof

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result in'earliar initiation of primary to secondary break flow--

accumu,lation.in the ruptured steam generator and earlier initiation of 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 would be overly n

conservative to include the simulation of turbine runback to

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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 South Texas reactor protection system, and turbine runback was simulated

- a,e 2.

Steam Generator Secondary Mass j

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initial secondary water mass in~the ruptured steam generator-

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was' determined by Reference 1 to be conservative for overfill.

As noted above, turbine runback was assumed to be initiated and was

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- s,The initial steam generator total fluid mass was conservatively t.

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AFW System Operation

,s For the reference plant analysis in WCAP-10698, reactor t g occurred on after the

- a,c SGTR.. and $1 was initiated on low pressierizer pressure at

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after reactor trip. The reactor and turbine trip and the assumed i

concurrent loss of offsite power will result in the termination of main feedwater flow and actuation of the AFW system. The SI signal

.c will also result in automatic isolation of the main feedwater system and actuation of the AFW system. The flow from the turbine-driven AFW 1

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-pumpLwill be available within approximately 10. seconds following the i

actuation signal, but the flow from tho' 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 turbina and motor-driven p sisinitiated{

he total AFW flow from all of the AFW.

I pueps was assumed to be distributed uniformly to each of the steam generators until operator actions are simulated to throttle AFW flow to control steam generator water level in accordance with the emergency procedures.

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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 51 initiatior., and it would not be conservativat'omo(e{the ]

Thus, for this analysis, the time of reactor trip i

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was determined by modeling the South Texas reactor protection system, andtgactuationoftheAFWsystemwasbasedonthe[

]Itwasassumedthatflowfromtheturbineandmotor-drivenAFW pumps is initiated at representativgtimedeIayfordeliveryofAFWflowtothesteam he maximum potential AFW flow rate of 675 gpm was used in the analysis for the turbine-driven pump and also for each of the motor-driven pumps.

C.

Operator Action Times

.In the event of an SGTR, the operator is required to take actions to g

stabilize the plant and terminate the primary to secondary leakage. The operator actions for SGTR recovery are provided in South Texas E0P L

PDP05-EO-E030 (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 2036v:1o/0e0789 7

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$..A.,A ofltheRCStorestoreinventory,andterminationofS1tostopprimaryto-secondary. leakage..These operator actions are described below.

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Identify the ruptured steam generator.

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- high' secondary side activity, as indicated by the main steamline radiation ~ monitors, steam generator blowdown radiation monitors, or condenser vacuum pump radiation monitor typically will provide the l

first indication of an SGTR event. The ruptured steam generator can be identified by an unexpected increase in steam generator level, or a high radiation indication on the corresponding main steamline radiation monitor or steam generator blowdown line radiation monitor, j

or high activity in any steam generator sample. For an SGTR that results in a reactor trip at high power as assumed in this analysis, the steam generator water level as indicated on the narrow range will 1

decrease significantly for all of the steam generators. The AFW flow wil'1 begin to refill the steam generators, distributing approximately equal flow to each of the steam generators. Since primary to secondary leakage adds additional liquid inventory to the ruptured steam generator, the water level in that steam generator will increase more rapidly. This response, as indicated by the steam generator j

water level instrumentation, provides confirmation of an SGTR event and also~ identifies the ruptured steam generator.-

J 2.

Isolate the ruptured steam generator from the int'act steam generators and isolate feedwater to the ruptured steam generator.

I 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, i

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 a

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between-the ruptured and intact steam generators as a necessary step:

toward-terminating primary to secondary, leakage.

In the South Texas E0F for steam generator tube rupture recovery, the operator is-directed to maintain the level in the ruptured steam generator between 5% and 50% on the narrow range instrument.

_Itwas assumed.thattherupturedsteamgenerator.willbeisolaledwhenlevel l

in the steam generator reaches midway between 5% and 50% or at 10 minutes, whichever is longer. Thus, for the South. Texas SGTR analysis, the ruptured. steam generator was assumed to be isolated at the time when the narrow range level reaches 27.5% or at 10 minutes, e

whichever was longer.

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

_ generators.

- Af ter 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 p

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 l

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.

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 L 2036v:1D/0907sp 9

u.-

i sufficient' reactor coolant subcooling and pressurizer inventory to mainta.in a reliable pressurizer level indication after SI flow is

' stopped. Since leakage from the primary side will continue after $1 J

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.

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The RCS depressurization is performed using normal pressurizer spray if the reactor coolant pumps (RCPs) are running, or auxiliary pressurizer' spray or the pressurizer PORVs if the RCPs are not running. Since offsite power is assumed to be lost at the time of l_

reactor trip, the RCPs are not running and thus normal pressurizer spray is not available.

In the South Texas SGTR recovery procedure, L

the first alternative is to use auxiliary pressurizer spray if normal l;

spray is not available, and the second alternative is to use a l

pressurizer PORV. Since the auxiliary pressurizer spray does not meet all of the requirements for safety grade equipment, credit would not normally be taken in the analysis for the use of auxiliary pressurizer spray and the analysis would be based on the use of a safety grade pressurizer-PORV. However, a scoping study has indicated that the use of the auxiliary spray produces more conservative results than the use of a PORV. Therefore, for this analysis, RCS depressurization was assumed to be performed using auxiliary pressurizer spray.

L The SGTR recovery procedure for South Texas instructs the operators te establish maximum charging flow after the RCS cooldown is completed but prior,to the RCS depressurization.

However, for RCS depressurization using the auxiliary spray system, the charging flow to the RCS must be isolated in order to utilize the auxiliary spray flow path to the pressurizer. Thus, it was assumed that the flow from two centrifugal charging pumps is supplied to the RCS, in addition to the flow from the SI pumps, for the time period from completion of the RCS cooldown until the initiation of RCS depressurization. For the 203sv:1o/090789 10

RCS depressurization,~it was assumed that'the normal charging fiow path is. isolated, and the auxiliary spray flow rate was conservatively based on the operation of only one charging pump.

It was also

.' conservatively assumed that the auxiliary spray flow rate is constant at the RCS pressure' corresponding to the beginning of the depressurization, whereas the spray flow rate would actually increase as the RCS pressure decreases. After the completion of-the RCS depressurization, it was assumed that the charging flow from two j

centrifugal charging pumps is retnitiated.

5.

Terminate SI to stop primary to secondary leakage.

t The previous actions will.have e'stablished 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, SI flow must be stopped to terminate primary to. secondary leakage. Primary to secondary leakage will continue after $1 flow is stopped untti RCS and ruptured steam generator pressures equalize.

Charging flow, letdown, and pressurizer heaters will then be controlled to prevent repressur12ation of.the-RCS and reinttiation of leakage into the ruptured steam generator. It was assumed that charging flow from the two centrifugal charging pumps continues for one minute following SI termination before the operators complete the action to eliminate excess charging flow.

L 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 s

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 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.

2036v:1D/091589 11

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- 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. Houston Lighting and Power Company has.

determined the corresponding operator action times to perform these l

. operations for South Texas. The operator actions and the corresponding _

operator action times used for the South Texas analysis are listed in Table 11.1.

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i 2036v:1D/0e0789 12

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r-1 te TABLE !!.1 STP SGTR ANALYSIS OPERATOR ACTION TIMES FOR DESIGN BASIS ANALYSIS 1

i Action =

Time (min)

' Identify and isolate ruptured SG 10 min or LOFTTR2' calculated time I

to reach 27.5% narrow range level in the ruptured SG, whichever is longer i

L perator action time to initiate 4

0

>cooldown Cooldown-Calculated by LOFTTR2 Operator action time to' initiate 3

depressurization Depressurization

' Calculated by LOFTTR2 Operator action time to initiate 2

.SI termination SI termination and pressure Calculated time sfter SI equalization termination for equalization of RCS and ruptured SG pressures, assuming excess charging flow from E

two centrifugal charging pumps for one minute after SI termination.

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Transient Deterintion The.LOFTTR2 analysis results for the margin'to overfill analysis are described below. The sequence of events for this transient is presented

' n Table 11.2.

i Following the tube rupture, reactor coolant flows from the primary into i

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 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 at 19 seconds 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 tne 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 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 AFM flow will be automatically initiated following reactor trip and the loss of offsite power.

The RCS pressure and pressurizar level continue to decrease after reactor trip as energy transfer to the secondary shrinks the reactor coolant and the tube rupture break flow continues to deplete primary inventory. When the RCS temperature differential begins to increase at approximately 35 seconds, the RCS pressure and pressurizer level decrease less rapidly.

The decrease in RCS inventory results in a low pressurizer pressure SI signal at 376 seconds. However, before the RCS pressure decreases to the 2036v:1D/090789 14

j 3

5

- shutoff head of the high head SI pumps, the pressurizer level goes offscale low. After the RCS pressure is below the shutoff head of the high head SI pumps..the SI flow rate maintains the reactor coolant inventory and the RCS pressure trends toward the equilibrium value where the SI flow rate equals the break' flow rate.

Since offsite power is assumed lost at reactor trip, the RCPs trip and a l

gradual transition to natural circulation flow occurs. Ismodiately following reactor trip the temperature differential across the core decreases as core power decays (see Figure II.4); however, the-temperature differential subsequently increases at approximately 35 seconds as the reactor coolant pumps coast down and natural circulation flow develops.

L The cold leg temperatures initially trend toward the steam generator temperature as the fluid residence time in the tube region increases. The RCS hot and' cold leg' temperatures then slowly decrease due to the continued addition of the auxiliary feedwater to the steam generators L

L until operator actions are initiated to control the auxiliary feedwater flow.

g MajorOperatorActions 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 isolating the auxiliary feedwater flow to the ruptured steam generator. As indicated previously, it is assumed that the ruptured steam generator will be identified and isolated when the narrow range level reaches 27.5% on the ruptured steam generator or at 10 minutes after initiation of the SGTR, whichever is longer. For the South Texas analysis, the time to reach 27.51 is less than 10 minutes, and thus it was assumed that the actions to isolate the ruptured steam generator are performed at 10 minutes. However, as noted previously, the limiting single failure was assumed to be g

2036v:10/090789 15

x op h'

]when the isolation is being performed. It was assumed that f

] fhus, the isolation of AFN flow to the ruptured steam generator was assumed to be completed at 12 minutes after the SGTR..The actual time used in_the analysis is 2 seconds longer because of the computer program numerical requirements for

' simulating the operator actions.

-2.-' Cool-Down the RCS to Establish Subcooling Margin e

After isolation of the ruptured steam generator is completed at 722 seconds,'a 4 minute operator action time is imposed prior to j

i initiating the cooldown. 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, the RCS is cooled by dumping igtactsteam steam to the atmosphere using the PORVs on the,be intact steam generators.

It was assumed that t

l generatorPORVsareopenedat962secondsfoitheRCScooldown. The

~

cooldown is continued unti1~ RCS subcooling at the ruptured steam j

generator. pressure is 20'F plus an allowance of 35'F for subcooling

= uncertainty. When these conditions are satisfied at 1410 seconds, it i

is assumed that the operator closes the intact steam generator PORVs to terminate the cooldown. This cooldown ensures that there will be l

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 f

the RCS temperature is shown in Figure II.4. As shown in Figure II.2, the RCS pressure also decreases initially during this cooldown process due to shrinkage of the reactor coolant, and then begins to increase due to the increased SI flow.

2036v:1D/090789 16

t 1

l6

- 3.

Depressurire RCS to Restore Inventory

.After the RCS cooldown, it is assumed that normal charging flow from two. centrifugal charging pumps is initiated. A 3 minute operator action time-is then included prior to the RCS depressurization. The RCS depressurization is performed to' assure adequate coolant inventory prior to terminating $1 flow. With the RCPs stopped, normal pressurizer spray is not available and thus the RCS is depressurized by using auxiliary pressurizer spray. The normal charging flow path if is isolated in order to uttitre the auxiliary spray flow path to the pressurizer. The RCS depressurization is initiated at 1600 seconds and continued untti any of the following conditions'are satisfied:

1 RCS pressure is less than the ruptured steam generator pressure and E

pressurizer level is greater than the allowance of 8% for pressurizer L

level uncertainty, or pressurizer level is greater than 701, or RCS 4

L subcooling is less than the 35'F allowance for subcooling uncertainty. For this case, the RCS depressurization is terminated

- due to high pressurizer level. The RCS depressurization reduces the break-flow as shown in Figure II.5, and increases SI flow to refill E

the pressurizer as shown in Figure II.I. After completion of the RCS

.depressurization, the charging flow from two centrifugal charging pumps was reinitiated.

4.

Terminate SI to Stop Primary to Secondary Leakage The previous actions have established adequate RCS subcooling, L

verified a secondary side heat sink, and restored the reactor coolant L

inventory to ensure that $1 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 35'F 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 increasing, and the pressurizer level is greater than the 81 allowance for uncertainty.

' 2036v:1D/0g0789 17 e

~,

xr r-e

--e um

-i-e.,

--en-emme.m.+.

-e.

+ -,-

,v,---

-s-.,---.

-e m

-.-*m-*-e

--.,,----v.,

,=,+-t e-.--,w,-

--ee-

{;j 4,

1 After depressurization-is completed, an operator action _ time of 2

]

minutes was assumed prior! to $1 termination. Since the above requirements are satisifed. $1-termination was performed at this j

time. The charging flow from two centrifugal charging pumps was continued from the end of the RCS depressurization until I minute after $1' termination, at which time it was assumed that excess charging flow is eliminated. After $1 termination and the elimination of excess charging flow, the RCS pressure begins to decrease as shown J

in Figure-II.2.

The intact steam generator PORVs also automatically open to dump steam to maintain the prescribed RCS temperature to ensure that subcooling is maintained. When the p0RVs are opened, the increased energy;

[,

transfer from primary to secondary also aids in the depressurization L

of the RCS to the ruptured steam generator pressure. The primary to

. secondary leakage continues after the SI flow and excess charging _ flow are Terminated until'the RCS and ruptured steam generator pressures H

equalize.

The primary to secondary break flow rate throughout the recovery operations is presented in Figure 1I.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 significantly less than 3

h the total steam generator volume of 7g83 ft. Therefore. it is L

concluded that overfill of the ruptured steam generator will not occur for a design basis SGTR for South Texas.

p

.;g_

s 2036v:1D/090789 18

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

$k.7

'I h.

TABLE 11.2.

STP SGTR ANALYSIS SEQUENCEOFEVENT}

NARGIN TO OVERFILL ANALYSIS l

t EVENT Time (sec)'

SG Tube Rupture 0

?n Reactor Trip 19 n

l

-SI Actuation'

'376

r L

Ruptured SG Is'o2ated 722 m

RCS Cooldown Initiated 962 L

RCS Cooldown. Terminated' 1410 Two Charging Pumps Started 1416 4

Charging Flow'to RCS Isolated 1600' u

.RCS Depressurization Initiated 1600 l

l RCS Depressurization Terminated 2072

'Two Charging Pumps Started 2074

~

SI Terminated' 2194 Excess Charging-Flow Eliminated 2256

-Steam Relief to Maintain RCS Subcooling 2674

Break Flow Toriinated 3786 2036v:10/090789 19

.g.

9 l

1 SOUTH-TEXAS STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERF]LL ANALYS1$

PRESSUR12ER LEVEL 90.'

te.'

l 72.<

I' b62.<

l r-a L

y d50.<

s-M E '0 '

l

(

k50.

ae.

l 18.'

8.B.

500. 1880. 15C2. 2000, 2500. 5880. 5582. 4800.

T!nt iStC1 i

l.

Figure 11.1 Pressurizer Level - Margin to Overfill Analysis, ao m a ntosasse 20 a

fi h-SOUTH TEXAS STEAM GENERATOR TU8E RUPTURE MARGIN TO OVERFILL ANALYSIS RCS PRESSURE 2500.<

l 2282.'

'2188.<

ij 2002..

L h1900.

c.

1600.'

1700.

j l

{1600.<

1580.-

j 1480.-

1582.<

i

'8.

582. 1980.1588. 2880. 2580. ESCO. 5580. 4880.

TIME ISCCl

-l i

figure 11.2 RCS Pressure - Nargin to Overtill Analysis

\\'

l L

k 2036v:10/082588 21

i l'

l p-l

)

L l

SOUTH TEXAS STEAM GENERATOR TUBE RUPTURE i

MAR 0!N TO OVERF]LL ANALYSIS SECONDARY PRESSURE

\\t 1580.-

l '

Ruptured SG 1

l 1200.

112D.1

.E fleee.

K'923.-

m-i E

, Sec..

78U

l y

Intact SGs see.

See.<

480 8.

583. 1988. 1580. 2888. 2588. 5880. 5580. 4888.

TINC (SCC)

Figure !!.3 Secondary Pressure - Nargin to Overfill Analysis.

ao m m osasse 22 l

I 1

e

-r.

r.--

1 i

I

.i i

SOUTH TEXAS STEAM DENERATOR TUBE RUPTURE MAR 0!N TO OVERFILL ANALYSIS

-)

INT ACT LOOP HDT AND COLD LEO RCS TEMPERATURES

]

sse.-

ses <

THOT-C

[

ysse.<

S" ' '

TCOLD 2

asc.<

?

4e0.-

sse..

'" 's.

see ises ises. asse. asse. sese. ssee. dese.

T1ttE isEC) l '

r Figure II.4 Intact Loop Hot and Cold Leg RCS Temperatures -

l-Norgin to Overfill Analysis so n v:1e ms sse 23

L; >

i e

't

?

t s

SOUTH TEXAS STEAM OENERATOR TUBE PUPTURE NARGIN TO OVERFILL ANALYSIS PRJMARY TO SECONDARY BREAK FLOW

?

55.<

sD.<

i.

ds.<

lj de.<

G ss.<

g

@se.<

B as <

d g re.<

5 s.'

i le.<

S.-

e.<

' ' 'e. see. sese. isse, asee ases, sese. ssee. 4sse.

TIME ISEC) i Figure 11.5 Primary to secondary Break Flow Rate -

Nargin to Overfill Analysis

' 3036v:10/082stt M

i J

i a

u i

l i

i SOUTH TEXAS STEAN GENERATOR TUBE RUPTURE MARDIN 10-DVERFILL ANALYS15 RUPTURED SG WATER VOLUME 7888.<

l-6588.;

6888.<

l 5588.<

a:D 8tt.<

E L

8 h

e 4588.-

n.88..

5588.<

I; l'

l 50ee'8.

588. 1888. 1588. 2888. 2588. 5888. 5588. 4888.

l TINC (SCCI 1

Figure 11.6 Ruptured SG Water Volume - Nargin to Overfill Analysis i

2$

3436v:1D482589

=.

k j

III. -ANALYSIS OF 0FFSITE RADIOLOGICAL CONSEQUENCES An~ analysis was also performed to determine the offsite radiological 0

' consequences for a design basis CT ~/ent for South Texas 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 3

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 atmosphere were calculated until break flow l

termination 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-calcula'ted results.

1.

Design Basis Accident The accident modeled is a double-ended break of one steam generator tubelocatedatthetopofthggubesheet]

thelocationofthebreak k

a,c.

However, as indicated subsequently, the breakflowflashingfractionwasconservativelycalculatedassuming l

that

~

'In addition, the iodine scrubbing effectiveness of the steam generator water was calculated based on the conservative assumption that the rupture is located near the top of the tube bundle ai the intersection of the outer tube row and the upper anti-vibration bar. The combination of these conservative assumptions regarding the l

break flow location results in a very conservative calculation of the offsite radiation doses.

It was also assumed that loss of offsite nosev:1o/oso7ss 26

. - ~ - -

h

.I J

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 l

reactor trip.

Based on the information in Reference 2, the most limiting single failure'with respect to of,fs,ite doses is

]tailureof{~

_g which will increase primary to secondary leakage and the mass release to the atmosphere. Pressure in the ruptured steam generator will remain belowthatin-theprimarysystemunti1{

]"ihus,fortheoffsitedoseanalysis,itwasassumed that the s, t.

2.

Conservative Assumptions L

Most of the conservative conditions and assumptions used for the L

margin to overfill analysis are also conservative for the offsite dose l

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, i

a.

Reactor Trip 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 J

the steam generator PORVs (and safety valves if their setpoints arereached). 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 South Texas reactor protection system, and this 2036cle/0007s9 27

.. ~...

n s

^

t timewasalsousedfortheoffsitedoseanalysis.[

J a., c.

1 4 -

b.- Steam Generator Secondary Mass If. steam generator overfill does not occur, a I a

L Jre,esults in a conservative prediction of offsite doses. Thus, for the offsite dose analysis, the initial secondary mass was assumed correspond to operation-a, t.

c,.AFW System Operation l

_ s, C.

In Reference 2, it was determined that results in an increase in the calculated offsite radiation doses for an:SGTR, whereas it was previously concluded that a, c.

is-conservativeforthemargin-tooverfillanalysIs.

However,itwasalsodemonstratedinReference2that]

L

~

ince the single failure assumed for the offsite radiation L

dose ~ana1

~

3 g is 1

Itisnotnecessarytoassumeanadditionalfailurein

~

the AFW system. Thus, each of the four AFW pumps were assumed to deliver flow to the associated steam generator, but a conservative minimum AFW flow of 500 gpm per pump was assumed for the offsite radiation dose analysis.

In addition, the delay tjag assumed for initiationofAFWflowwas{

since this assumption results in a conservative calculation of the mass releases for the offsite radiation dose analysis.

' 2036v:1o420789 28

. i I

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

~

nce 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

]

hus the ass g tion that

~

{

is g

. conservative for the SGTR analysis.

3.

Operator Action Times The major operator actions required for the recovery from an SGTR are i

discussed in Section II.C and the operator action times used for the overfill an'alytis are presented in Table. II.1. The operator action L

times in Table II.1 were also used for the offsite dose analysis.

However, for the offsite dose analysis, the

[thetiIdetherupturedsteam

~

generator is isolated.

It was assumed that the operators a,e before proceeding with the subsequent recovery operations.

~~

The

-4C fiouston Lightingand'PowerCompanyhasdeterminedthatanoperatorcan]

]ihus,itwasassumedthatthe{

e, e.

] a,t.

an additional After the delaytimeof4 minutes (Table 11.1)wasassumedfor.theoperator action time to initiate the RCS cooldown.

i

~ 2036v:1o/oeO789 29 m-

.u-..-..

. ~.

j m

+

4.

Tranttent beterint1on 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 overft11 analysis until the time when the ruptured steam generator is isolated. Thetransientbehaylorisdifferentafterthistimss{nge it is assumed that v6en

~

the isolation is performed.

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 at 19 seconds on an l

L overtemperature delta-T signal. After reactor trip, core power rapidly decreases to decay heat levels and the RCS depressurization continues. 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 PORV setpoint as shown in Figure III.2.

Pressurizer level also continues to decrease following reactor trip as shown in Figure III.3. When the RCS temperature differential begins to increase at approximately 35 seconds-(see Figures III.4 and III.5) as'the reactor coolant pumps coast down and natural circulation flow develops, the RCS pressure and pressure level decrease less rapidly.

The decreasing pressurizer pressure leads to an automatic SI signal on low pressurtzer pressure at 463 seconds.

However, before the RCS l

pressure decreases to the shutoff head of the high head SI pumps, the L

pressurizer level goes offscale low. After the RCS pressure is below

'the shutoff head of the high head SI pumps, the SI flow rate maintains the reactor coolant inventory and the RCS pressure decrease is reversed.

2036v:1D/091589 30

1

I Major Operator Actions-1

-1

1. _ Identify and Isolate the Ruptured Steam Generator As indicated in Table II.1, it is assumed that the ruptured steam generator will be identified and isolated at 10 minutes after the initiation of the SGTR or when the narrow range level reaches 27.5%, whichever time-is longer. Since the time to reach 27.5%

narrow range level is slightly greater than 10 minutes, it was assumed that the actions to isolate the ruptured steam generator are performed at this time. The

~ Y this time. 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 L

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 L

decrease. as shown in Figure III.5, until the AFW flow is L

controlled to maintain the specified level in the intact steam generators. These effe:ts result in a further decrease in the RCS pressure. However, when the RCS pressure decreases below the shutoff head of the high head SI pumps, the SI flow slows the rate of pressure decrease and subsequently causes the RCS pressure to increase again.

It is assumed-that the time required for the operator'toidentifythatthe{

~fs15

- minutes'. Thus, the isolation of the ruptured steam" generator is cospleted at 1586 seconds and the depressurization of ruptured steam generator is terminated. At this time, the ruptured steam a

generator pressure begins to increase to the PORV setpoint and the primary to secondary break flow begins to decrease. Because the SI flow rate exceeds the break flow rate, the rate of RCS repressurization increases.

~ 203sv:1D/oe07a9 31

. - ~..

4

-_j'

,e

2.. Cool Down the RCS to Establish Subcooling.Nargin a,L Afterthe{

u a-4' minuteoperatoractiontimeisimposedpriortoinitiationof cooldown. The depressurization of the ruptured steam generator.

effects the RCS cooldown target temperature since the temperature is dependent upon the pressure in:the ruptured steam generator.

(

Since offsite-power is lost, the RCS is tooled by dumping steam to the atmosphere using the intact steam get. orator PORVs. The' cooldown is continued until RCS subcooling at the ruptured steam.

=

generator pressure is 20'F plus an allowance of 35'F for l

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 L

the time for cooldown - The cooldown is initiated at 1826 seconds-

{

and is completed at 2842 seconds.

(

The reduction in the intact steam generator pressures required to l.

accomplish ~the cooldown is shown in Figure III.2, and the effect l.

of the cooldown on the RCS temperature is shown in Figure 111.5.

L The RCS pressure also decreases during this cooldown process due to shrinkage of the reactor coolant as shown in Figure 111.1.

. 3.- Depressurize to Restore Inventory i

After the RCS cooldown, it is assumed that normal charging flow from two centrifugal-charging pumps.is initiated. A 3 minute i

operator action time is then included prior to the RCS depressurization. The RCS is depressurized to assure adequate

.c L

coolant inventory prior to terminating $1 flow. With the RCPs j

stopped, normal pressurizer spray is not available and thus the l.

RCS is depressurized by using auxiliary pressurizer spray. The normal charging flow path is isolated in order to utilize the l

2036v:1o/090789 32

..,4

ff j

W l

auxiliary spray flow path to the pressuriser. The RCS depressurization is initiated at 3024. seconds and continued until any of the following conditions are satisfied:

RCS pressure is j

less than the ruptured steam generator pressure and pressurizer

^

level is greater than the allowance of 85 for pressurizer level uncertainty, or pressurizer level is greater than 70%, or RCS l

subcooling is less than the 35'F allowance for subcooling uncertainty. For this case, the RCS depressurization is l

terminaten due to high pressuriser level. The RCS depressurization reduces the break flow as shown in Figure III.7, 1

and increases $1 flow to refill the pressuriser as shown in Figure l -

!!!.3. After completion of the RCS depressurization, the charging l

flow from two centrifugal charging pumps was reinitiated.

L l

l 4.

Terminata SI to Stop Primary to Secondary Leakage l

L l-The previous actions have established adequate RCS subcooling, verified a secondary side heat sink, and restored the reactor coolant inventory to ensure that $1 flow is no longer needed, i

When these actions have been completed, the 51 flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary leakage. The 51 flow is terminated at this time if RCS subcooling is greater than the 35'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 increasing, and tne pressuriser level is greater than the 8%

i allowance for uncertair.ty.

L After depressurization is completed, an operator action time of 2 minutes was assumed prior to S1 termination. Since the above requirements are satisfied, si termination is performed at this l

time. The charging flow from two centrifugal charging pumps was continued from the end of RCS deptessurization until 1 minute after $1 termination, at which time it was assumed that 203sv;1D/0e07:s 33 9

e

,-......,.m._2

_,,_w.m w,,mm.,__

_,._ mms.

..,,,,m

,_.._,.,,m

_. ~

i i

I l

emcess charging flow is eliminated. After 51 termination and the elimination of excess charging flow, the RCS pressure decreases as shown in figure !!!.1. The differential pressure between the RCS and the ruptured steam generator is shown in Figure !!!.6. Figure

!!!.7 shows that the primary to secondary leakage continues af ter j

'the $1 flow and excess charging flow are stopped until the RCS and j

ruptured steam generator pressures equalize.

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

i For this case, the water volume in the ruptured steam generator is I

substantially less than the total steam generator volume of 7983 ft 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 !!!.9.

i i

e' I

S r

P no m.tomsores 34

t.

j TABLE !!!.1 STP SGTR ANALYS!$

stoutuv. Dr nynnis 0FFSITE RADIATJDN DJ5H. ANALYSIS EvtWT TIME (sec)

SG Tube Rupture 0

i

[

Reactor Trip 19

+

SI Actuation 463 l

i i

Ruptured SG lsolated 680 I

a,c 684 a, c 1586 i

l t

RCS Cooldown Initiated 1826

\\

L l

RCS Cooldown Terminated 2842

}

l l

Two Charging Pumps Started 2842 Charging Flow to RCS Isolated 3024 RCS Depressurization Initiated 3024 RCS Depressurization Terminated 3424 f

- Two Charging Pumps-Started 3426 51 Terminated 3546 Excess Charging Flow Eliminated 360g Break Flow Terminated 4854 monvas/oeores 35

)

}

l t

J souTN TEXAS ETEAM CENERATOR TUBE RUPTURE l

OFFSITE DOSE ANALYS!$

I RCS PRES $URE i

Ps...<

c

.i....

...D.<

E 19...-

C ine. <

l giv.e..

pi....

is..

is.e.-

l is....

L

' 'a..

11MC istCl t

l Figure !!!.1 RCS Pressure - Offsite Radiation Dose Analysis souv.iessasse 36

l l

'i l

t j

k SOUTH TEXAS STEAM GENERATOR TUBE RUPTURC OFFSITE DOSE ANALYSIS SECONDARY PRESSURE

.l Inet.<

Intact SGs t

stet..

l

)

Esete..

l 0

(

{eee.<

r IsB8.<

j l

l W 4ee.< Ruptured SG i

L 298..

i 8 *S.

1888.

3988.

s000.

4988.

5888.

1 TIME istC) l^

h Figure !!!.2 Secondary Pressure - Offsite Radiation Dose Analysis 38Mv;1D/042ste 37

~.. ~

SOUTH TEXAS STEAM DENERATOR TUSE RUPTURE OFFSITE DOSE ANALYS15 PRESSUR12ER LEVEL 188.'

M..

to.<

l p

78.<

h60.'.

M g

St..

48.<

r 58.'

30.<

le.<

'O.

1800.

2008.

5880.

4888.

6888.

TIME ISEC) 1 Figure 111.3 Pressuriser Level - Offsite Radiation Dose Analysis 20h10/DB2649 38

,y J

r

)

1 SOUTH TEXAS STEAM GENERATOR TUSE RUPTURE DFFSITE DOSE ANALYSIS RUPTURED LOOP HOT AND COLD LEO RCS TEMPERATURES t

sse.<

1 see.,

TH0T

{

C u

Esse.<

b a

l o 500.<

r k

TCOLD A

THOT E ase.<

g...

e sse..

TCOLD

's.

1998.

3888.

sees.

4888.

5888.

TsnC isCC1 Figure 111.4 Ruptured Loop Het and Cold Leg RCS Temperatures :

Offsite Radiation Dose Analysis 39 aoskto/osasse

J I

4 i

i

}

l j

J l

SOUTH TEXAS STEAM GENERATOR TUBE RUPTURE D'FSITE DOSE ANALYS1$

INTACT LOOP HOT AND COLD LEO RCS TEMPERATURES l

658.<

THOT see.,

t C

$ESB.'

f l

i -.

l Ste.<

t W

E TCOLD

,,d50.-

l E

L g488.<

W 550.<

0.

1998.

2006.-

5808.

4988.

5098.

TIMC ISCCI l

I Figure !!!.5 Intact Loop Hot and Cold Le RCS Temperatures -

Offsite Radiation Dose Anal sis asasvao/estses 40

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

i i

SOUTH TEXAS STEAM GENERATOR TURE RUPTURE l

DFFSITE DOSE ANALYSIS f

D)FFERENTIAL PRESSURE BETWEEN RCS AND RUPTURED SG I788.<

r f

1989.'

[

(DO.<

t i

E-608.'

i e 428.'

f

,88..

8.'

'8.

1988.

3088.

Se88.

4988.

5808.

j

~

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

Offsite Radiation Dose Analysis som:10/ passes -

41

s.

e

(.

p.

SOUTH TEXAS STE AM DENERATOR TUSE RUPTURE OFFS 3TE DOSE ANALY$15 PRIMARY TO SECONDARY OREAK FLOW

[

SS.'

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

Offsite Radiation Dose Analysis 1

sossanneases 42

l i

i i

I i

I SOUTH TEXAS STEAM GENERATOR TUBE RUPTURE I

^

OFFS!TE DOSE ANALYS!$

RUPTURED SG WATER VOLUME 4568.'

eMt. <

l f

4988.<

l 6758.'

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TINC (SCC) s Figure !!!.8 Ruptured SG Water Volume - Offsite Radiation I

Dose Analysis l

1 l

l asa r. w es Sas 43

[

l i

1 1

I SOUTH TEXAS STEAM GENERATOR TUBE RUPTURE DFFSITE DOSE ANALYSIS RUPTURED SG WATER MASS f

t i

P48888.<

j 278888.<

t

$288828.'

l l

m I

w It0200.'

t W

E N 18:008.'

e 148888.'

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1998.

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4988.

5000.

11MC 15CCI Figure !!!.9 Ruptured SG Water Nass - Offsite Radiation Dose Analysis zon a nssasse 44

i 5.

Mass Releases The mass releases were determined for use in evaluating the exclusion ares 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the accident and from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the accident. The releases for-1 0-2 hours are used to calculate the radiation doses at the exclusion area boundary for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exposure, and the releases for 0-8 hours j

are used to calculate the radiation doses at the low population zone for the duration of the accident, t

In the LOFTTR2 analyses, the $6TR recovery actions in South-Texas E0P PDP05-E0-E030 (E-3) were simulated until the termination of primary to secondary leakage. After the primary to secondary leakage is j

terminated, the operators will continue the $6TR recovery actions to prepare the plant for cooldown to cold shutdown conditinns. When these recovery actions are completed, the plant should be cooled and depressurized to cold shutdown conditions, was assumed that the cooldewn is performed using South Texas E0P PDP05-EO-ES33 (ES-3.3), POST-SGTR C00LDOWN U$1NG STEAN DUNP, since this method results in a conservative l

evaluation of the long term mass releases for the offsite dose analysis.

The high le' val actions for the the post-56TR cooldown method using steam dump in South Texas E0P P0P05-EO-ES33 (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.

2036do/o00789 -

45

s 4

s i

Thess include isolating the cold leg $1 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.

CoolDownRCStoResidualHeatRemoval(RHR)SystemTemperature i

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 l -

prcierred means for cooling the RCS is steam dump to the condenser i

since this minimizes the radiological releases and conserves i

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 cooldown is performed using steam dump to the atmosphere via the intact steam generator PORVs. When the RHR system operating temperature is reached, the cooldown is stopped o

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

3.

Depressurize RCS to RHR System Pressure When the cooldown to RHR system temperature is completed, the pressure in the ruptured steam generator is decreated 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 PORV 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 depressurized by releasing steam via the PORV. As the ruptured steam generator 20m:1D/o007s9 46

- ~. -

i pressure is reduced, the RCS pressure is maintained equal to the pressure in the ruptured steam gen 6rator in order to prevent in-leakage of secondary side water or additional primary to secondary leakage. Although normal pressurizer spray is the l

preferred asans of RCS pressure control, auxiliary spray or a l

pressuriser PORY can be used to control RCS pressure if j

pressurizer spray is not available.

I i

4.

Cool Down to Cold Shutdown j

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

)

i p

The methodology in Reference 2 was used to calculate the mass releases for the South Texas analysis. The methodology and the results of the L

calculations are discussed below, a.

Nothodology for Calculation of Nass Releases 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 I

steam generators, and the primary to secondary leakage into the

(

~

ruptured steam generator were determined from the LOFTTR2 results for the period from the ihitiation of the accident until the leakage is terminated.

l Following the termination of leakage, it was assumed that the RCS and intact steam gnerator 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 47 20:4v:1D/000789

1 down the RCS to the RNR system operating toaperature of 350*F at the maximum allowable cooldown rate of 100'F/hr. The RCS and the i

intact steam generator temperatures at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months were then determined

-ag, steamreleasesandthefeedwaterflowsfortheintactsteh generator for the period from leakage termination until 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months were determined from

~Mncetherupturedsteam 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 continusd after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months until the RNR system in-service temperature of 350*F is reached, j

L Depressurization of the ruptured steam generator was then assumed to be performed inmediately following the completion of the RCS j

cooldown.

The ruptured steam generator was assumed to be j

depressurized to the RNR in-service pressure of 350 psia via steam release from the ruptured steam generator PORV, 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 RNR operating conditions are t

L completed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months were determined for the intact steam generator from

~

b steam released L

t from the ruptured steam generator from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months was determined based on n, t.

'2036v:10/031583 48 l

~

After 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , it is assumed that further plant cooldown to' cold i

shutdown as well as long-term cooling is provided by the RNR

)

system. Therefore, the steam releases to the atmosphere are terminated after RNA in-service conditions are assumed to be

{

reached at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , j

b.

Mass Release Results l

f The mass release calculations were per, formed using the methodology l

discussed above. For the time period from initiation of the l

accident until leakage termination, the releases were determined from the LOFTTR2 results for the time prior to reactor trip and l

following reactor trip. Since the condenser is in service untti j

reactor trip, any radioactivity released to the atmosphere prior to reactor trip will be through the condenser vacuum pump exhaust. After reactor trip, the releases to the atmosphere are assumed to be via the steam generator PORVs. The mass release rates to the atmosphere from the LOFTTR2 analysis are presented in Figures !!!.10 and !!!.11 for the ruptured and intact steam generators, respectively, for the time period until leakage j

termination, j

The mass releases calculated from the time of leakage termination until 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and from 2-8 hours are also assumed to be released to the ata) sphere via the steam generator PORVs. The mass releases for the SGTR event for each of the time intervals i

considered are presented in Table III.2.

The most releases prior to break flow termination, from break flow termination until 2 l

hours, and from 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months are summarized in Table III.3. The results indicate that approximately 12g,300 lbs of steam are released from the ruptured steam generator to the atmosphere in L

the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . A total of 186,000 lbs of primary water is j

transferred to the secondary side of the ruptured steam generator l'

before the break flow is terminated.

2036v:10/0g158g 4g

l i

TABLE !!!.2

$TP $GTR ANALYS!$

WA$$ RELEASES OFF$1TE RADIATION DOSE ANALY$l$

TOTAL MAS $TLOW (POUNDS)

TIME PERIOD i

1 0-TRIP TRIP -

TMSEP -

TTBRK -

T2HR$ -

l TMSEP TTBRK T2 HRS TRHR 1

Ruptured SG t

Condenser 23,000 0

0 0

0 t

Atmosphere 0

20,400 108.900 0

41,700 Feedwater 22,000 49,200 3,300 0

0 Intact SGs Condenser 68,400 0

0 0

0 l

Atmosphere 0

57,300 285,800 228,900 1,051,100 l

Foodwater 68,400 149,700 483,600 243,500 1,063,400 Break Flow 900 24,800 160,300 0

0 TRIP

= Time of reactor trip = 19 sec.

i TNSEP = Time when water reaches the moisture separators = 631 sec.

TTBRK = Time when break flow is terminated = 4854 sec.

T2HR$ = Time at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months = 7200 sec.

[

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

i

)

sonao/conse 50

~,~&

_...-.y..

_______._______________,m

i t

TABLE !!!.3 STP $6TR ANALYS!$.

$lM4ARIZE0 NASS RELPSES i

0FFSITE RADIATION DOSE ANALYS!$

t TOTALI4ASSFLOW(POUNDS) i 0-TTBRK -

2 HRS -

l TTSRK 2 HRS 8 HRS Ruptured SG Condenser 23,000 0

0 t

l Atoosphere 129,300 0

41,700 Feedwater 74,500 0

0 t

l 1

Intact SGs t

n Condenser 68,400 0

0 L

Atmosphere 343,100 228,900 1,051.100 l

l 1

i Feedwater 701,700 243,500-1,063.400 l

(

l l

l 4

Break Flow 186,000 0

0 l

I i

l l

1 2036c1D/0eO7ee 51 I

't 4

SOUTH TEXAS STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS RUPTURED SG ATNOSPHERIC MASS RELEASES l

663.<

U k

I 5 588

l f498.<

I e

hf588..

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E l

l -.

e 188.'

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l 0.g,

gggg, 3333, goes.

dest.

5868.

11ME ISEC) i Figure !!!.10 Ruptured SG Wass Release Rate to the Atacsphere -

Offsite Radiation Dose Analysis 30 W 3S/983589 N

_m..

e, I

I i

k-I i

i SOUTH TEXAS STE AM OENERATOR TUSE RUPTURE j

OFFSITE DOSE ANALYSIS t

INT ACT SOS Ain0$PHERIC MASS RELE ASE

- i 8

2008.<

t l

O 1968.<

k i

ei....

{

1498.<

i W

lite.<

I u 1000.'

w 900.'

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Figure 111.11 Intact SGs Mass Release Rete to the Atmosphere -

Offsite Radiation Dose Analysis l

l sonaomnes 53 w

y -,+

,y,--,,..,,c--y--,,=~.3,~,-

,,#.m,,we-

-3.w.,,

.,....,+.em m

a

l 1

S.

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 Techn1 cal Spec 1f1 cation 11mit for primary coolant attivity and primary to secondary leakage for suffletent 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 PORVs and safety valves and via the condenser

)

vacuum pump exhaust.

The quantity of radioactivity releas'd to the environment, due to an SGTR, l

e depends upon primary and secondary coolant activity, todine spiking i

effects, primary to secondary break flow, break flow flashing fractions.

[

attenuation of todine carried by the flashed portion of the break flow, l

partitioning of todine between the 11guld and steam phases, the mass of fluid released from the generator and 11guld-vapor partitioning in the F

turbine condenser hot well. All of these parameters were conservatively evaluated in a manner consistent with the recommendations of Standard i

Review Plan Section 15.6.3.

I t

1.

Og11gn Ratin Analvtical Ant - tiemt The major assumptions and parameters used in the analysis are iteatred l

in Table !!!.4.

j 2.

Eaurce Term Calculattans The radionuclide concentrations in the primary and secondary system,

[

prior to and following the SGTR are determined as follows:

2036v:10/03158g 54

i a.

The iodine concentrations in the reactor coolant will be based l

upon preaccident and accident initiated iodine spikes.

i 1.

Accident Initiated Spike - The initial primary coolant iodine concentration is 1 vCi/gm of Dose Equivalent (D.E.) 1-131.

Following the primary system depressurination associated with the SGTR, an iodine spike is initiated in the primary system i

which increases the lodine 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 l

concentration.

The initial appearance rate can'be written as follows:

i j

Pg=Ag g

i where:

Pg = equilibrium appearance rate for iodine nuclide i j

Ag = equilibrium RCS inventory of iodine nuclide i' corresponding to 1 vCi/gm of D.E. 1-131 l

kg = removal coefficient for iodine nuclide i

_ s,e, The duration of the spike,{

is sufficient to

],4 increase the initial RCS 1-131 inventory by a factor of 1

11. Preaccident Spike - A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine i

concentration from 1 to 60 pCi/ gram of D.E. 1-131.

b.

The initial secondary coolant iodine concentration is D.1 vCi/ gras, of D.E.1-131.

a03s d o/0907a0 55

~-wn er-r-,,.

e~e.

s,a-n,.--,.<.n-a

..w.m.,

1 l

c.

The chemical form of todine in the primary and secondary coolant j

is assumed to be elemental.-

d.

The initial noble pas concentrations in the reactor coolant are based upon 11, fuel defects. These concentrations were taken from Table 15.A-2 of the South Texas FSAR.

i 3.

Date calculattans The lodine transport model utilized in this analysis was proposed by I

postma and Tam (Reference 4). The model considers break flow flashing, droplet size, bubble scrubbing, steaming, and partitioning.

The model assumes that a fractio'n of the lodine carried by the break flow becomes airborne immediately due to flashing and atomization.

Removal credit is taken for scrubbing of todine contained in the atostred coolant droplets when the rupture site is below the secondary water level.

The fraction of primary coolant todine which is not assumed to become atrberne tunedtately mixes with the secondary water i

and is assumed to become airborne at a rate proportional to the steaming rate and the iodine partition coefficient.

This analysis conservatively assumes an todine partition coefficient of 100 between the steam generator 11guld and steam phases. The model takes no scrubbing credit when the rupture site is above the secondary water 1evel. Droplet removal by the dryers is assumed to be negligible.

l The iodine transport model is illustrated in Figure III.12.

l 1

l The following assumptions and parameters were used to calculate the activity released to the atmosphere and the offstte 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.

2036v:1D/09158g 56

'e l

b.

The time dependent fraction of rupture flow that flashes to steam

)

and is immediately released to the environment is presented in Figure 111.13, c.

In the iodine transport model, the time dependent iodine removal

]

efficiency for scrubbing of steam bubbles as they rise from the j

rupture site to the water surface conservatively assumes that the j

rupture is located at the insorsection of the outer tube row and the upper anti-vibration bar. However, in accordance with the l

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 collapsed water level relative to the top of j

the tubes in the ruptured and intact steam generators is shown in i

Figure !!!.14.

The iodine scrubbing efficiency is determined by l

themethodsuggestedbyPostmaandTam(Ref.4). However, since i

the collapsed water leval in the ruptured steam generator is below the rupture site for most of the time when the rupture flow is flashing, the effect of iodine scrubbing is very small aiid has been conservatively neglected for this analysis, i

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

A "E IA Il~'II) r j

j j

where:

total iodine released to the environment by l

-A a

r flashed primary coolant IA)

(integrated activity in rupture flow during time

=

interval j) (flashing fraction for time interval j) eff) iodine scrubbing efficiency during time interval j

=

2036cloMe07at 57

.m..

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

i' j

(

d.

The total primary to secondary leak rate is assumed to be 1.0 spe ass allowed by the Technical Specifications. The leak rate is assumed to be 0.70 ppm for the three intact steam generators and I

0.3 ppm for the ruptured steam generator. The leakage to the intact steam generators is assumed to persist for the duration of

.the accident.

l e.

The iodine partition coefficient between the liquid and ste6m of the ruptured and intact steen generators is asswed to be 100.

l f.

No credit was taken for radioactive decay during release and i

transport, or for cloud depletion by ground deposition during l

transport to the site boundary or outr,r boundary of the low population zone.

l i

g.

Short-termatmosphericdispersionfactors(x/Qs)foraccident analysis and breathing rates are provided in Table !!!.8. The f

breathing rates were obtained from NRC Regulatory Guide 1.4, (Ref.

5).

1 l-I k

I P

l'-

t to3sr1D/De07:9 58

.... ~...

. _ ~

i

)

1 4.

Offsite bote calculation

]

Offsite thyroid doses are calculated using the equation:

l i

t (IAR)gj (BR)j (a/Q)j kh E

where integrated activity of iodine nuclide i released (IAR),)

during the time interval j in Ci' breathing rate during time interval j in (BR))

=

3 i

meter /second (Table III.8) atmospheric dispersion factor during time interval (x/0))

=

i jinseconds/ meter 3 (Table !!!.8) thyroid dose conversion factor via inhalation for (DCF)q

=

iodine nuclide i in res/Ci (Table !!!.9) thyroid dose via inhalation in tem D

Th Offsite whole-body gamma doses are calculated using the equation:

I,g (IAR)g3(a/Q))

D. 0.25 y

No credit is taken for cloud depletion by ground deposition or by L

radioactive decay during transport to the exclusion area boundary or to the outer boundary of the low-population zone, l

l 2036v:1D/0g2089 59 l

t

I l

l i

where:

integrated activity of noble gas or iodine j

(IAR)y

=

nuclide i released during time interval j in ci

atmospheric dispersion factor during time

)

J/

(x/Q))

=

3 interval j in seconds /m I

average gamma energy for noble gas or iodine g

nuclide i in Nev/ dis (Table III 10) 1 D

whole body gamma dose due to immersion in rem-y

=

Offsite beta-skin doses are calculated using the equation:

I Dg = 0.23 i

(IAR)g) (x/Q))

gg L

l l

where:

integrated activity of noble gas or iodine (IAR)g).

nuclide i released during time interval j in Ci

i atmospheric dispersion factor during time (x/Q))

=

3 intervaljinseconds/m E

average beta energy for noble gas or iodine gg i

nuclide i in Nev/ dis (Table 111.10) beta-skin dose due to innersion in res 0

8 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.

2036v:1D/091589 60

{q:lw 8 p

b pp y

' lb :

,+

1 5.

Results?

Thyroid, whole-body game, and beta-skin doses at the Exclusion Area Boundary and the outer boundary of the Low Population Zone are:

presented in Table III.11. All' doses are within the allowable-guidelines as specified by_ Standard Review Plan 15.6,3 and 10CFR100.

l l

1 b

-t L

i...

I p-l' I

~2036v:fo/000789 61 i

l 40

'O' 1

y, s

l y

n d

7

(.

l TABLE !!I.4 f

i STP SGTR ANALYSIS.

l l

~

PARAMETERS USED IN EVALUATING

. t.

. RADIOLOGICAL-CONSEQUENCES 1

I.

Source Data'

(,

j A.

Core power level, NWt 4100

)

B.. Total steam generator tube 1.0 leakage, prior to accident, spa J

C.- Reactor. coolant. activity:

l 1.

Accident Initiated Spike The initial RC iodine ll q

activities based on-1 j

pCi/ gram of..D.E. 1-131 F

are presented in 1

L Table !!!.5. The iodine appearance rates assumed 7

for the accident L

initiated spike-are j

presented in Table III.6.

1 h

J j-

'2.

Preaccident Spike Priaary coolant lodine activities based on 60 pCi/ gram of D.E. 1-131 are presented in Table III.S.

3.

Noble Gas Activity The initial RC noble gas i

activities based on 1%

fuel defects are presented in Table III.7.

' 203sv:1o/000789 62

. ~

i i

s' i

L TABLEIII.4(Sheet 2)

D.

Secondary system initial activity -

Dose equivalent of 0.1

-yCi/gm'of I-131, presented in Table III.S.

l E.

Reactor coolant mass, grams 2.6 x'108 7

F.

Initial Steam generator water mass 4.9 x 10

^

(each), grams i

d

~ G.

Offsite power Lost at time of reactor trip H.

Primary-to-secondary leakage 8

duration for intact SG, hrs.

I.

Species of iodine 100 percent elemental J

II.

Activity Release Data A.

Ruptured steam generator 1.

Rupture flow See Table 111.2 J

2.

Rupture flow flashing fraction See Figure III.13.

3.

Iodine scrubbing efficiency Negligible j

4.

Total steam release,1bs See Table III.2 5.

Iodine partition coefficient 100 mossv:1omeo7ss 63

a'3 E

E e

TABLE !!!.4 (Sheet 3).

I

. ~.

6.

Location of tube rupture Intersection of. outer tube row and upper I

anti-vibration bar i

l-B.

Intact steam generators 1.

Total primary-to-secondary 0.7 leakage, gpe' -

L

2. ' Total steam release. Ibs See Table 111.2 1

3.

Iodine partition coefficient 100 t

C.

Condenser g

L

'1.

Iodine partition coefficient 10D p

D.

Atmospheric Dispersion Factors See Table 111.8-b L

l:

p.

l.

?

o 1

i r

. 303Sv:1D/0007a0 64

4 3

l l

. TABLE !!!.5.

l STP SGTR ANALYSIS, 10 DINE SPECIFIC ACTIVITIES.

IN THE PRIMARY AND SECONDARY COOLANT BASED ON 1. 60 AND 0.1 vC1/ oram DF D.E.1-131*

J Specific Activity (vCi/ne)

Primary Coolant Secondary Coolant-Nuclide 1 vC1/om 60 vCi/cm 0.1 vCi/am j

I I-131 0.75 45.0 0.075 1-132 0.88 52.8 0.088

- i

'l-133 1.19 71.4 0.120 1-134 0.18 10.8 0.018 1-135 0.66' 39.6 0,066

Consistent with the STP Technical. Specifications.

h 2036v:1D/000'89 65 1

'me-t u.

..,.-,,,.,_...-.-w.-

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

,.-e--

,e

-]

-F

7.

-1 i

TABLE III.6' STP SGTR ANALYS!$

l 10 DINE SPIKE APPEARANCE RATES (CURIES /SECOND)

-I-131 1-132 1-133 1-134 I-135 2.2

'12.1 4.8 5.7 4.4 i

b p

Y t

T i i mosev:10/oso7se -

66

3

<t 4

.j 1

i e

m

'i TABLE !!!.7' STP SGTR ANALYS!$

NOBLE GAS SPECIFIC ACTIVITIES IN THE:

REACTOR COOLANT BASED ON ll, FUEL DEFECTS Nuclide Specifie' Activity-(vCi/cm)

)

'Xe-131m 2.0

'l Xe-133m 16.0 1

e l

Xe-133 250.0 Xe-135m 0.46

'Xe-135 6.8 Xe-138 0.64 l

Kr-85m 2.0 Kr 7.3 1

Kr-87 1.2 Kr-88 3.6 i

O l

I s

l_

i' aossv:1o/osorse 67

. ~.

.)

. g '

i TABLE !!!.8 STP SGTR ANALYSIS ATMOSPHERIC DISPERSION FACTOR $ AND BREATHING RATES 4

Time Exclusion Area Boundary Low Population Breathing 3

3 3

(hours) x/0 (Sec/m )

'Zonex/0(Sec/m1 Rate (m /See) [5] 2 1.3 x 10'4 3.8 x 10 3.47 x 10'4

-5 1.6 x 10 3.47 x 10'4

-5 2-8 l

l-I-

l l

l 2036r.1D/090780 60

.~

'll{i e

m;

y 1

TABLE-!!!.9 STP SGTR ANALYSIS, l

THYROID DOSE CONVERSION FACTORS

-(Rem / Curie)[Ref.6)-

Nuclide I-131-

'1.49 x 106 j

4 I-132 1.43 x 10-5 I-133 2.69 x 10 3

('

I-134 3.73 x 10 a

li f

4 I

I-135 5.60 x 10 1

V l'

j.

I

~

L L

Lc 1'

i --

l

, aossem m otas 69

c 3

L i

L

+

i k

TA8LE III.10 i.

i STP 1GTR ANALYSIS AVERAGE GAMA AND BETA ENERGY FOR MOBLE GARES AND 1EINES (Nev/ dis) [Ref. 7).

j Nuclide l

Eg y

-f Xe-131m 0.0029 0.16 l

i

~Xe-133m 0.02 0.212 Xe-133 0.03 0.153 Xe-135m 0.43 0.099 t

Xe-135 0.246-0.325 i

Xe-138 1.2 0.66 i

Kr-85m 0.156

~0.253 Kr-85 0.0023 0.251 Kr-87' O.793 1.33 Kr-88 2.21

'0.248 I-131 0.38 0.19 I-132 2.2 0.52

'I-133 0.6 0.42 I-134 2.6 0.69 l

I-135 1.4 0.43 l

1 2036v:ID/091589 70

m,v i

t-F TABLE III.11 STP SCTR ANALYSIS OFFSIT( RADIATION DOSES Doses (Rem)

Calculatod-

~ Allowable Value fuideline Value thef. A1 1.

Accident Initiated Indine Snike Exclusion Area Bound &ry (0-2 hr.)

Thyroid Dose 4.0 30 Whole - Body Gamma Dose 0.067 2.5*

' Beta - Skin Dose 0.110-2.5*

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

Thyroid Dose 1.2 30 Whole - Body Gamma Dose 0.020 2.5*

Beta'- Skin Dose 0.033 2.5*

3

- 2 '. Pro-Acc1dont'Indine Snike Exclusion Area'8oundary (0-2-hr.)

Thyrold Dose 15.6 300 Nhole'- Body Gamma Dose 0.069-25*

i I

Beta --Skin Dose 0.110 25*

Low Population' Zone (0-8 hr.)

l Thyroid Dose 4.6 300 l

Whole - Body Gesum Dose 0.020 25*

Seta - Skin Dose 0.034 25*

1

Assumed to apply to the sum of the whole-body gamma and beta-skin doses.

l.

h 2036v:10/091589 71

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aw i.,,t

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"k Figure !!!.12 Iodine Transport Model - Offsite i

Radiation Dose Analysis t

n.

i:

- 20Mv:10/082849 72 I

7 f

ee -.. -.

.--,--.._.-...-4_

_-__,_______.-______-.-_._--.....__,--,.-m

. - ~ - - -... -

,-_e

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y.

4

(

J

'l SOUTH TEXAS STEAM GENERATOR TUBE RUPTURE 0FFSITE DOSE ANALYSIS BREAK FLOW FLASHING FRACT]ON

.tS<

f-

.14' El.12<

V.,.

7

.88 4

~

w.66-W.

f f.w.

.82-

's.

1980.

3988.

.5888.

4688.

6888.

TIME ISCC)

L l=

Figure !!!.13 Break Flow Flashing Fraction - Offsite Radiation Dose Analysis 1

2036c1D/0825ts 73

1~

fh 8

R

-l S0llTH TEMAS STEan GENERATOR TUDE RUPTURE R

I:3 0FFSITE DOSE ANALYSIS ~

SG-SECONDARY LEVEL ABOVE TOP OF TUDES I

p

'130.'

g IM. '

l f"*'

Intact ~SGs L

I as.'

as.

l-7

.a W ' e, l:

. ti l88' 30.

l Ruptured SG u

g Se.

l,

s,

3000.

.Sete.

400s.

Stee.

TIME IRC) i l-l l

t f.

1~

l.

=

(

Figure 111.14 SG Water Level Above Top of Tubes -

L Offsite Radiation Dose Analysis 1;

E L-messe.iossasse 74 D

1 L

l P

1 k

IV. CONCLUS10N-u An evaluation has been performed for a design basis SGTR' event for the South l-Texas Units 1 and 2 to demonstrate that the potential consequences are L

. acceptable. 'An analysis was performed to demonstrate margin to steam L

generator overfill assuming the limiting single failure relative-to overfill.

The limiting single failure is the failure of{

~

Yhe results of this

' analysis indicate that f.he recovery actions can be coupleted 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 thisanalysis,itwasassumedthat[

..g primary to secondary break' flow and the mass relaases 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 South-Texas would be acceptable.

203sc1D/090709 75

.~

Y

. ERRATA SHEET FOR WCAP-12370 (NON-PROPRIETARY) 1 I

'V.

REFERENCES-

!. Lewis LHuang. Behnke, Fittante,LGelman; "SGTR Analysis Nethodology_to 1

. Determine the Wargin to Steam Generator.Overf111." NCAP-10698-P-A er

[ PROPRIETARY]/NCAP-10750-A (NON-PROPRIETARY], August 1987.

2., Lewis, Huang.. Rubin, " Evaluation of Offd e Radiation Doses for a' Steam g

Generator Tube Rupture Accident," Supplement I to NCAP-10698-P-A

-[ PROPRIETARY)/ Supplement I to NCAP-10750-A [NON-PROPRIETARY), March 1986.

3.

Lewis, Muang.. Rubin,'Nurray, Roidt, Hopkins, " Evaluation of Steam

' Generator Overf111 Due to a Steam Generator Tube Rupture Accident,"

NCAP-11002 [ PROPRIETARY)/NCAP-11003 [NON-PROPRIETARY), February 1986.

i-4; Postma, A. K., Tam, P. S., " Iodine Behavior in a PNR 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 Water Reactors," June 1974.

6. :NRC Regulatory Guide 1.109, Rev.~1, " Calculation of Annual Doses to Wan From Routine Releases of Reactor Effluents for the Purpose of Evaluating

-Compliance with 10 CFR Part 50 Appendix I," October 1977.

7.

Bell, W.

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

)

ORNL-4628, 1973.

8.- Standard Review Plan, Section 15.6.3, " Radiological Consequences of Steam Generator Tube Failure," NUREG-0800, July 1981.

2036v:1D/092589 76

c.

]

7 V.

REFERENCES

1. : Lewis, Huang, 8ehnke, Fittante, Gelman, "SGTR Analysis Nethodology to Determine the Margin to Steam Generator Overfill," WCAP-10750-A, August 1987.

2.

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

l l

3.

Lewis, Huang, Rubin, Nurray, Reidt, Hopkins, " Evaluation of Steam L

Generator Overfill Due to a Steam Generator Tube Rupture Accident,"

l WCAP-11003, February 1986.

Postma, A. K., Tam, P. S., " Iodine Behavior in a PWR Coolir.g System 4.

Following a Postulated Steam Generator Tube Rupture", NUREG-0409.

1 L

5.

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

6.

NRC Regulatory Guide 1.109, Rev. 1 " Calculation of Annual Doses to Man L

'From Routine Releases of Reactor Effluents for the Purpose of Evaluating L

Compliance with 10 CFR Part 50 Appendix I", October 1977.

7.

8 ell, M. S. "0RIGEN - The ORNL !sotope Generation and Depletion Code",

DRNL-8628, 1973.

l 8.

Standard Review' Plan, Section 15.6.3, " Radiological Consequences of Steam

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

l l

p.

I ao3sv:1D/0007s9 76 a -

- -.

ML19325D122

Text

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