ML20151V886

LOFTTR2 Analysis for Steam Generator Tube Rupture for Diablo Canyon Power Plant Units 1 & 2

ML20151V886
Person / Time
Site:	Diablo Canyon
Issue date:	02/29/1988
From:	Holderbaum D, Robert Lewis, Miller T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML16341E628	List: ML20151V886
References
WCAP-11724, NUDOCS 8805030228
Download: ML20151V886 (81)

Text

. _ _ _ _ _ _

WESTINGHOUSE CLASS 3 WCAP-11724 LOFTTR2 ANALYSIS FOR A ,

STEAM GENERATOR TUBE RUPTURE FOR THE DIABLO CANYON POWER PLANT UNITS 1 AND 2 D. F. Holderbaum R. N. Lewis T. A. Miller K. Rubin FEBRUARY 1988 Nuclear Safety Department Westinghouse Electric Corporation  :

Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 (c) 1988 by Westinghouse Electric Corporation 11s7c10/021ssa 8805030228 880429 .

PDR ADOCK 05000275 P _ _. , D.CD )

TABLE OF CONTENTS

.P. age 1

la INTRODUCTION i

- 11. ANALYSIS OF MARGIN TO STEAM GENERATOR OVERFILL 4 f

l l

l A. Design Basis Accident 4 B. Conservative Assumptions 5 C. Operator Action Times 8 D. Transient Description 13  :

III. ANALYSIS OF OFFSITE RADIOLOGICAL CONSEQUENCES 25 A. Thermal and Hydraulic Analysis 25

1. Design Basis Accident 25

2. Conservative Assumptions 26

3. Operator Action Times 28

4. Transient Description 29

5. Mass Releases 43 B. Offsite Radiation Oose Analysis 52 IV. CONCLUSION 74 V. REFERENCES 75 11s7v:1o/C21988 i

LIST OF TABLES Table Title Page 11.1 Operator Action Times for Design Basis Analysis 12 II.2 Sequence of Events - Margin to Overfill Analysis 18 111.1 Sequence of Events - Offsite Radiation Dose Analysis 33 III.2 Mass Releases - Offsite Radiation Dose Analysis 48 III.3 Sumarized Mass Releases - Offsite Radiation Dose 49 Analysis III.4 Parameters Used in Evaluating Radiological 60 Consequences III.5 lodine Specific Activities in the Primary and Secondary 63 Coolant III.6 lodine Spike Appearance Rates 64 III.7 Noble Gas Specific Activities in th. R n ivr Coolant 65 Based on 1% Fuel Defects .

III.8 Atmospheric Dispersion Factors and Breathing Rates 66 III.9 Thyroid Dose Conversion Factors 67 111.10 Average Gama Energy for Noble Gases 68 111.11 Offsite Radiation Doses 69 1157v;10/021988 ii

l l

l LIST OF FIGURES Figure Title .Page 11.1 Pressurizer level - Margin to Overfi ll Ana'ysis 19 II.2 RCS Pressure - Margin to Overfill'Ana' lysis 20 11.3 Secondary Pressure - Margin to Overfill Analysis 21 II.4 Intact Loco Hot and Cold leg RCS Temperatures - 22 Margin to Overfill Analysis 81.5 Primary to Secondary Break Flow Rate - Margin to 23 Overfill Analysis 81.6 Ruptured SG Water Volume - Margin to Overfill Analysis 24 fl!.1 RCS Pressure - Offsite Radiation Dose Analysis 34 811.2 Secondary Pressure - Offsite Radiation Dose Analysis 35 III.3 Pressurizer Level - Offsite Radiation Dose Analysis 36 III.4 Ruptured Loop Hot and Cold leg RCS Temperatures - 37 Offsite Radiation Dose Analysis 811.5 Intact Loop Hot and Cold Leg RCS Temperatures - 38 Offsite Radiation Dose Analysis 811.6 Differential Pressure Between RCS and Ruptured 39 SG - Offsite Radiation Dose Analysis i III 7 Primary to Secondary Break Flow Rate - Offsite 40 Radiation Dose Analysis l

1157tlo/021988 iii

l LIST OF FIGURES (Cont)

Figure Titie P_aSe III.8 Ruptured SG Water Volume - Offsite Radiation Dose 41 Analysis III.9 Ruptured SG Water Mass - Offsite Radiation Dose Analysis 42

\

III.10 Ruptured SG Mass Release Rate to the Atmosphere - 50 Offsite Radiation Oose Analysis III.11 Intact SGs Mass Release Rate to the Atmosphere - 51 Offsite Radiation Dese Analysis l l

111.12 Iodine Transport Model - Offsite Radiation Dose Analysis 70 I!!.13 Break Flow Flashing Fraction - Offsite Radiation 71 Dose Analysis 111.14 SG Water Level Above Top of Tubes - Offsite 72 Radiation Dose Ar,slysis 111.15 Iodine Scrubbing Efficiency - Offsite Radiation Dose 73 Analysis 11s7v;10/021988 iV

l I. INTRODUCTION t .

An evaluation for a design basis steam generator tube rupture (SGTR) event has been performed for the Diablo Canyon Power Plant (DCPP), 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 tha' the calculated offsite radiation doses are less than the allowable guidelines.

• The Diablo Canyon Power Plant employs two essentially identical Westinghouse pressurized water reactor (PWR) units. The reactor coolant system for each unit has four reactor coolant loops with Series 51 steam generators. While the reactors, structures, and all auxiliary equipment are substantially identical for the two units, there is a difference in the capability of the turbine generators. Consequently, the licensed power ratings are 3338 MWt for Unit 1 and 3411 MWt for Unit 2. However, the SGTR evaluation has been performed using the limiting parameters for either Unit 1 or Unit 2 so that the analysis is applicable for both units. It is also noted that both units are currently operating with Westinghouse standard fuel, but it is planned to install optimized fuel in the future. Therefore, the most limiting parameters for standard or optimized fuel were also used for the SGTR evaluation such that the results are applicable for operation with either fuel type.

The steam generator tube rupture analyses were performed for Diablo Canyon using the methodology developed in WCAP-10750 (Reference 1) and Supplement 1 to WCAP-10750 (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. 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 analysis mathodology includes the simulation of the operator actions for recovery from a steam generator tube rupture based on the Diablo Canyon Emergency Procedures, which were developed from the Westinghouse Owners Group 11s7elo/021988 1

Emergency Response Guidelines (ERGS). In subsequent references to the Diablo Canyon Emergency Procedures, it should be noted that the designation for the Emergency Procedures is the same as the corresponding Westinghouse Owners Group ERGS.

An SGTR results in the leakage of contaminated reactor coolant into the secondary system and subsequent release of a por' tion of the activity to the atmosphere. Theref' ore, 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 tesoverfill. 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.

For the margin to overfill analysis, it was assumed that the e LOFTTR2 analysistodeterminethemargintooverfillwasperformedfortIetimeperiod 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 d'emonstrate that overfill does not occur. The results of this analysis demonstrate that there is margin to steam generator overfill for Diablo Canyon.

Since steam generator overfill.does not occur, the results of the offsite radiation dose analysis represent the limiting consequences for Diablo I Canyon. For the analysis of the offsite radiation doses, ae Theprimarytosecondarybreakflow i

andthesteamreleasestotheatmosphlrefromboththerupturedandintact 1157v:1o/021ssa 2

l steam generators were calculated for use in determining the activity released to the atmosphere. The mass releases were calculated with the LOFTTR2 program from the initiation of the event until termination of the break f. low. For the time period following break flow termination, steam releases from and feedwater flows to the ruptured and intact steam generators were determined from a mass and energy balance using the calculated RCS and steam generator conditions at the time of leakage termination. 'The mass release information was used to calculate the radiation doses at the exclusion area boundary and low population zone assuming that the primary coolant activity is at the maximum allowable Technical Specification limit prior to the accident. The results of this analysis show that the offsite doses for Diablo Canyon are within the allowable guidelines specified in the Standard Review Plan, NUREG-0800, Section 15.6.3, and 10CFR100, s

i 1157v;1o/02198: 3

II. ANALYSIS OF MARGIN TO STEAM GENERATOR OVERFILL An analysis was performed te determine the margin to steam generator overfill for a design basis SGTR event for Diablo Canyon. 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 sequtnce 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 located at the t -

of the tube sheet

'f$e location of the break -

- c.

s'It was alsoassumedthatlossofoffsitepoweroccursatthetimeofrUctor trip, and the worst rod was assumed to be stuck at the operating position at reactor trip.

The most limiting single failure with respect to steam generator overfill wasdeterminedtobe] _

,however, a sensitivity study indicates that the most limiting singie' failure for the four-loop Diablo Canyon plants is

- , s, s.

The Diablo Canyon AFW system consists of two motor-driven pumps, and one turbine-driven pump with a capacity equal to the combined capacity of the two motor-driven pumps. Each motor-driven pump normally feeds two steam 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 from 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 1157v:1o/021988 4

control inventory in the intact steam generators. - 4,c requires the operator to perform additional actions to AL Itwasdetermined that a failurea,s of ~

is slightly more limiting relative to overfill than a failure- c of a,Since the a s,s - -

for this case, the

- n,s Thus, it was assumed that the as inaccordance

~ "

with Reference 1, it was assumed that

-- n,s --

This a,L -

results in additional primary to secondary leakage as well as

-ae which decreases the margin to steam generator overfill.

B. Conservative Assumotions Sensitivity studies were performed previously to identify the initial 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 assumotions which were used in Reference 1 were also used in ths LOFTTR2 analysis to determine the margin to steam generator overfill for D'ablo Canyon with the exception of the following differences.

1157v:10/0219as 5

1. Reactor Trip and Turbine Runback.

A turbine runback can either be initiated automatically or the operator can manually reduce the turbine load following an SGTR to attempt to prevent a reactor trip. For the reference plant analysis in WCAP-10750,sreactor trip was calculated to occur at approximatelya,e s,d turbine runback to an was

~

simulatedEasedonarunbackrateoY~ Ihe effect of turbine runback was conservatively simulated by -

Nowever,ifreactortrip 4,,

occurs prior to t rbine runback to would not be possible. It is noted that earlier reactor trip will result in earlier initiation of primary to secondary break flow accumulation in the ruptured steam generator and earlier initiation of 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 ti:ne after the SGTR occurs rather than at a fixed time after reactor trip. It would be overly

, ~

conservative to include the simulation of turbine runback to 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 Diablo Canyon reactor protection system, and turbine runback was simulated- a, t. -

2. Steam Generator Secondary Mass

- e, L ,

initial secondary water mass in the ruptured steam generator A

waIdetEmined by Reference 1 to be conservative for overfill. As noted above, turbine runback was assumed to be initiated and was simula,ted by

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

- a, c.

em.

1157v.l o/021988 6

3. AFW System Operation For the reference plant analysis in WCAP-10750 reactor trig occurred on{ '$fterthe

~ ~

SGTR,andSIwasinitiatedonlowpressurizerpressureIt

~

after reactor trip. ThereactorandturbinetripandtheIssumed concurrent loss of offsite power will result in the termination of -

main feedwater flow and actuation of the AFW system. The SI signal will also result in automatic isolation of the main feedwater syster and actuation of the AFW system. The flow from the turbine-driven 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 pu,mps is initiated

'The total AFW flow from all of the AFW pumpswasassumedtobedistributeduniformlytoeachofthesteam 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 ec thepressureat

~

the time of reactor trip may be significantly higher than the SI t initiation setpoint. In this event, there may be a significant time delay between reactor trip and SI initiation, and it would not be

~

conservativetomodglthe g

~

ihus,forthisanalysis,thetimeofreactortrip was determined by modeling the Diablo Canyon reactor protection system,andygetuat.ionoftheAFWsystemwasbasedonthe{

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

]fo'rdeliveryofAFWflowto l

1157v:1o/021988 7

I the steam generators. The total'AFW flow assumed for the analysis is the combined capability of the turbine-driven pump and both motor-driven pumps.

C. Operator Action Times In the event of an SGTR, the operator is req'uired to take actions to stabilize the plant and terminate the primary to secondary leakage. The operator actions for SGTR recovery are provided in Diablo Canyon Emergency Procedure E-3, and these actions were explicitly modeled in this analysis. The operator actions modeled include identification and isolation of the ruptured steam generator, cooldown and depressurization of the RCS to restore inventory, and termination of SI to stop primary to secondary leakage. These operator actions are described below.

1. Identify the ruptured steam generator.

High secondary side activity, as indicated by the main steamline radiation monitors, the air ejector radiation monitor, or steam ,

generator blowdown radiation monitor, typically will provide the first indication of an SGTR event. The ruptured steam generator can be i identified by an unexpected increase in steam generator level, or a high radiation indication on the corresponding main steamline radiation monitor or from a radiation survey of the main eteamlines.

For an SGTR that results in a reactor trip at high power as assumed in this analysis, the steam generator water level will decrease off-scale -

on the narrow range for all of the 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 liquid inventory to ta ruptured steam generator, the water level will return to the narrow range earlier in that steam generator and will continue to increase more  !

i rapidly. This response, as indicated by the steam generator water level instrumentation, provides confirmation of an SGTR event and also l

identifies the ruptured steam generator.

11s7v;1o/0219ss 8

l

.l

2. Isolate the ruptur.ed steam generator from the intact 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, this also reduces the possibility of ov rfilling the ruptured steam generator with water by 1) minimizing the accumulation of feedwater flow and 2) enabling th. operator to establish a pressure differential between the ruptured and intact steam generators as a necessary step toward terminating primary to secondary leakage. In the Diablo Canyon Emergency Procedure for steam generator tube rupture recovery, the operator is directed to maintain the level in the ruptured steam generator between 4% and 33% on the narrcw range instrument. However,

- e, e it was assumed that the ruptured steam generator will be isolatedwdnlevelinthesteamgeneratorreachesmidwaybetween4%

and 50% or at 10 minutes, whichever is longer. Thus, for the Diablo Canyon analysis, the ruptured steam generator was assumed to be isolated at 27 percent narrow range level or at 10 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 j rapidly as possible to less than the saturation temperature i corresponding to the ruptured steam generator pressure by dumping steam from only the intact' steam generators. This ensures adequate i 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 i

1157v.1D/021ss: 9 1

---.-,,n,w,- , -

---

y w--,- - - , -e g----.----my---- --,,m ,

l l

l offsit.e power is assumed to be lost at reactor trip for this analysis, j 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 flen'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 subcocling 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 pressurizar level remains on span. The "excess" amount required i depends on Rr.S 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

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

leakage. Primary to secondary leakage will continue af ter SI flow is l stopped until RCS and ruptured steam generator pressures equalize.

l l

l 11s7v:1o/02198: 10 ,

l l

l 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.

Since these major recovery actions are modeled in the SGTR analysis, it is necessary +o 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 st'eps. 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 eflected by the time required for the operator to perform the intermediate action steps. I s,s The operator actions and the corresponding operator action times are listed in Table 11.1.

1-l t

i i

1157v:1o/02198: 11

l TABLE II.1 l DCPP SGTR ANALYSIS 1 OPERATOR ACTION TIMES FOR DESIGN BASIS ANALYSIS Action .

Time (min)

Identify and isolate ruptured SG- 10 , min or LOFTTR2 calculated time to reach 27% narrow range level in the ruptured SG, whichever is longer Operator action time to initiate 5 cooldown Cooldown Calculated by LOFTTR2 Operator action time to initiate 4 depressurizetion Depressurization Calculated by LOFTTR2 Operator action time to initiate 1 SI termination l

SI termination and pressure Calculated time for SI termination equalization and equalization of RCS and ruptured SG pressures f

l 1

1157v:1o/0219ss 12

De Transient Description The LOFTTR2 analysis results for the margin to overfill analysis are described below. The sequence of events for this transient is presented in Table II.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 11.1. The RCS pressure also decreases as shown in Figure II.2 as

- tae steam bubble in the pressurizer expands. As the RCS pressure decreases due to the continued primary to secondary leakage, automatic reactor trio 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 tho' turbine is terminated. The steam dump system is desigm to actuats following reactor trip to limit the increase in secondary pressure, but the steam dump valves remain closed due to the loss of condenser vacuum reculting 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 11.3. The main feedwater flow will be l terminated and AFW flow will be automatically initiated following reactor trip and the loss of offsite pcwor.

The RCS pressure decreases more rapidly after reactor trip as energy '

transfer to the secondary shrinks the reactor coolant and the tube rupture bieak 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 maintains the reacter coolant inventory ,

4 1

1157v:1o/021ssa 13 I

.. - - _ - - . . .- . - . - - - _ , _ - - - . - - - _ _ = - -

4 and the pressurizer level begins to stabilize. The RCS pressure aisc 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 gradual transition to natural circulation flow occurs. Immediately following reactor trip the tempeiature diffe'rential across the core l

decreases as core power decays (see Figure II.4), however, the temperature differential subsequently increases as natural circulation flow develops. t The cold leg terporatures trend toward the. steam generator temperature as }

the fluid residence time in the tube region increases. The RCS hot leg temperatures slowly decrease due to the continued addition of the auxiliary feedwater to the steam generators until operator actions are initiated to cool down the RCS. ,

i Major Operator Actions +

l

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 l indicated previously, it is assumed that the ruptured steam genurator ,

will be identified and isolated when the narrow range level reiches  !

27% on the ruptured steam generator or at 10 minutes af ter initiation j

) of the SGTR, whichever is longer. For the Diablo Canyon anaissis, the l 4

1 time to reach 27% is less than 10 minutes, and thus it was assumed  ;

that *he actions to isolate the ruptured steam generator are performed at 10 minutes. However, as noted previously, the limiting single ,

, failure was assumed to '

-AL i

witentheisolationisbeingperformed. It was assumed that

- ~

l

=

= Ajd ,

1 l

l I

l 1157tio/021988 14 i  !

i

AeC Thus, the isolation of the ruptured steam generator was assumedIobecompletedat12minutesaftertheSGTR;

2. Cool down the RCS to Establish Subcooling Margin After isolation of the ruptured steam generator is completed at 12 minutes, a 5 minute operator action time is imposed prior to initiating the cooldown. The actual delay time used in the analysis is 4 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, the RCS is cooled by dumping steam to the atmosphere using the PORVs a , e.

on the intact steam generators. It was assumed that the  ;

inta;t steam generatcr PORVs are opened at 1024 seconds for the RCS cocidown. The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20'F plus an allowance of 20'F for subcooling uncertainty, When these conditions are satisfied at 1614 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 acecmplish the cooldown is shown in Figure II.3, and the effect of the cooldown on the RCS temperature is shown in Figure II.4. The RCS pressure aise 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 4 minute operator action time is included

- prior to depressurization. The RCS depressurization is initiated at 1858 seconds to assure adaquate coolant inventory prior to terminating 11sh:1o/0219:s 15

S1 flow. With the RCPs stopped, normal pressurizer spray is not available and thus the RCS is depressurized by openirg a pressurizer PORV. The depressurization is continued until any 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 4% for pressurizer level uncertainty, or pressurizer

^

level is greater than 77%, or RCS subcooling is less than the 20'F allowance for subcooling uncertainty. The RCS depressurization reduces the break flow as shown in Figure II.5, and increases SI flow to rafill the pressurizer as shown in Figure 11.1.

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 20'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 4% allowance for uncertainty.  ;

To assure that the RCS pressure 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 1 minute was assumed prior to SI termination. Since the above requirements are satisifed, SI termination was performed at this time. After 51 termination, the RCS pressure begins to decrease as shown 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 PORVs are opened, the increased energy transfer from primary to secondary 1157tio/021988 16

also aids in the depressurization of the RCS to the ruptured stsam generator pressure. The primary to secondary leakage continues after the SI flow is terminated until the RCS and ruptured steam generator pressures equalize. ,

The primary to secondary break flow rato throughout the recovery operations is presented in Figure II.S. The water volume in the ruptured steam generator is presented as a function of time in Figure 11.6. It is noted that the water volume in the ruptured steam generator when the break flow is terminated is less than the total steam generator volume of 5759 ft3 . Therefore, it is concluded that overfill of the ruptured steam generator will not occur for a design basis SGTR for Diablo Canyon.

l l

I e

1157v.io/o21ssa 17

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t- TABLE II.2 '

. 3.1 P ..

DCPP SGTR ANALYSIS- . a, r

., .. . z: : -

4 SE0VENCE OF EVENTS lw d s;.m j MARGIN TO OVERFILL. ANALYSIS Tyf JQ]

. y ., , ,

c.;el . EVENT Time (see) g,,,

,......m,

.l

a. 4- 4 :.. ; .

g t.

SG Tube Rupture O syc; .

4y. y -

m g.m

_ 'b Reactor Trip 121  %.D.:..

,; ' 'k[k

SI Actuated 150

.nq mq l Ruptured SG Isolated 720 . : .

3 tim . . . .

RCS Cooldown Initiated 1024 g', . % y g;;

.-'- il@'.}

h RCS Cooldown Terminated 1614 y-(

...+.

r .-.

. j..- RCS Depressurization Initiated 1858 ^[..A.j

.c . . . .

i;:  : ' ,.

@l RCS Depressurization Terminated 2002 ,.j ;#.}

n,- ..

9.,..,..

.) - SI Terminated 2062 'hi

k. --.

ep:

1. 1 - .

g@Q

-1 Steam Relief to Maintain RCS Subcooling 2518

.-i '

t. 7

.,'4 J>

E Break Flow Terminated 2728  :.,4 f[.,

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. . ., .. ,a ., - .. w , , . . . . . m .w s.n , , s. , , ~

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS PRESSURIZER LEVEL 60.

63.

.G.

J 50.

W b

$ 40.

2 3

tjsc.

c.

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' O.

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l TIME ISEC) l Figure !!.1 Pressurizer Level - Margin to Overfill Analysis 1157v.te/czises 19

. L .,

E

'} .

KI -

.,.b

, ~ . . .

. 4 .. .

..+v.

rn c0

. .. . , 3 .y;

. 4 . .' : .

1 DIABLO CANYON STEAM GENERATOR TUBE RUPTURE < ; ;

MARGIN TO OVERFILL ANALYS15 ..

"a-RCS PRESSURE c- 4 '.V. y r ,..,-

t~

. j6'3 i_ 2403. o u

• I T.'

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, 1000'O. 1000. 2000. 5000. 4000.

TIME ISEC)

Y '

K L"

t Figure 11.2 RCS Pressure - Margin to Overfill Analysis r

=

1157v:10:',21988 20

~

r zl k ~

..% .,e.I -

.. n.

' 4'4 ' ,

' . ,' .;,4l DIABLO CANYON STEAM GENERATOR TUBE RUPTURE . :- ,

,.-,,4 -

y MARGIN TO OVERFILL ANALYSIS - -

SECONDARY PRESSURE .DS .

Y3

'I 1100.

.'Qh h- RUPIVRID SG

~ w; V

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TIMC (SCCI v

i

- Figure 11.3 Secondary Pressure - Margin to Overfill Analysis E

u sn.ieregissa 21 r-

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.- 1 , ,'

__... g_

33

! .-,, .e . ,

w : _ ,9,< ..

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DIABLO CANYON STEAM GENERATOR TUBE RUPTURE

i4 .+

, ~ '

MARGIN TO OVERFILL ANALYSIS ' , , ' - },

vr.. _ ~

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INTACT LOOP HOT AND COLO LEG RCS TEMPERATURES '

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TIME (SEC1 Figure 11.4 Intact Loop Hot and Cold Leg RCS Temperatures -  ;

Margin to Overfill Analysis B

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS PRIMARY TO SECONDARY BREAK FLOW 93.

60.-\

72.

_ 60.

U d

s 50, 5

$ 43.

d hI !2.

t G

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Figure !!.5 Primary to Secondary Break Flow Rate - -

Margin to Overfill Analysis its7tio/czissa 23

f- T i

OIABLO CANYON STEAM GENERATOR TUBE RUPTURE MARGIN TO OVERFILL ANALYSIS RUPTURED SG WATER VOLUME 6000..

- .- - -. -- **.Lt 2.".**.i*&2 1.*1""_ ,

ssee.

G 5 scre..

> 4523.<

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TIME ISCCI Figure !!.6 Ruptured SG Water Volume - Margin to Overfill Analysis I

1157v.10/021ssa 24

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

The thermal and hydraulic and the offsite radiation dose analyses were performed using the methodology developed in References 1 and J.

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

i

1. Design Basis Accident The accident modeled is a double-ended break of one steam generator tube located at the top of the tube sheet ~ ~

~

ihe location of the break

- a , C.

However, as indicated subsequently, the breakflowflashingfractionwasconservativelycalculatedassuming that

~ ~

Inaddition,theiodinescrubbingeffectivenessofthe

~

steam generator water was calculated based on the conservative assumption that the rupture is located near the top of the tube bundle

- at the intersection of the outer tube row and the upper anti-vibration bar. The combination of these conservative assumptions regarding the break flow location results in a very censervative calculation of the offsite radiation doses. It was also assumed that less of offsite 11s7v:1o/0219sa 25

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

Based on the information in Reference 2, the most limiting single failure with respect to offsite doses is ~ ~

' Failure of

~ " "

ch willincreaseprimarytosecondaryleakageandthemassreleIsetothe atmosphere. Pressure in the ruptured steam generator will remain below that in the primary system until hus, for the offsite dose analysis, it was assumed that the

- a, e.

2. Conservative Assumotions l

Most of the conservative conditions and assumptions used for the i margin to overfill analysis are also conservative for the offsite dose analysis, and thus most of the same assumptions were used for both analyses. The major differences in the assumptions which were used for the LOFTTR2 analysis for offsite doses are discussed below,

a. Reactor Trio and Turbine Runback An earlier reactor trip is conservative for the offsite dose analysis, similar to the case for the overfill analysis. Que to the assumed loss of offsite power, the condenser is not available for steam releases once the reactor is tripped. Consequently, after reactor trip, steam is released to the atmosphere through the steam generator 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 modeling the Diablo Canyon reactor protection system, and this ushiorcatssa 26 i

1 time was also used for the offsite dose analysis.

- , a ,e.

b. Steam Generator Secondary Mass '

~

Ifsteamgeneratoroverfilldoesngtoccur,a

, N'esultsinaconservative prediction of offsite doses. Yhus,fortheoffsitedoseanalysis, the initial secondary mass was assumed to correspond to operation

- s,L

c. AFW System Operation

- ~

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

doses for an GTR, whereas it was previously concluded that

- a, iscoiservativeforthemargintooverfillanalysIs.

However, it was also demonstrated in Reference 2 that '

AE hince the single failure assumed for the offsite radiation

~

dose analysis is a,e -

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 initiation of AFW ,

flow was

~

6,L

)

its7ticionissa 27

d. Flashing Fraction ,

When calculating the amount of break flow that flashes to steam,

- a e.

Sincethetube rupture flow actually consists of flow from the hot leg and cold leg sides of the steam generator, the temperature of the combined

~ #

flow will be -. ae ihustheassumptionthat

- s,a is conservative for the SGTR analysis.

I

3. Operater Action Timej 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 in Table 11.1 were also used for the offsite dcse analysis.

However, for the offsite dose analysis, the ~

f. the tine the ruptured steam

~

generator is isolated. It was assumed that the operators 6,e before proceeding with the subsequent recovery operations.

~ ~

The

]a,c Pacific Gas and Electric Company has determined that an operator can' -.

~ ~

h us, it was assumed that the

~

ae dfterthe

~

$ltadditionaldelaytimeof 5 minutes (TahleII.1)wasassumedfortFIoperatoractiontimeto initiale the RCS cooldown.

11s7<1 orc 21ssa 28

I

4. Tr'ansient Description 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 th's ruptilred steam generator is f isolated. The transient ~

behavior is different after this time ~since it is assumed that the

~ ~

Nn .

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 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 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 !!!.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, theRCSpressureandpressurizerleveltendtostabilizeuntil[ = s,C i ,

Major Operator Actions

1. Identify and Isolate the Ruptured Steam Generator

As indicated in Table II.1, it is assumed that the ruptured steam j generator will be identified and isolated at 10 minutes after the v

!- initiation of the SGTR or when the narrow range level reaches 27%,

l whichever time is longer. Since the time to reach 27% narrow range level is slightly greater than 10 minutes, it was assumed

~

that the actions to isolate the ruptured steam generator are 4

11s7c10/czissa 29

,y.,-, , .-_-.--....r-. g , , . , , - , _ . - , . , - _ - . , . _ , , -- , ...-,-m. , _ . _ - _. - - -

~

l performedatthisting.g The atthisIlme. 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 decrease in the ruptured loop temper'atures as shown in Figure

!!I.4. The intact steam generator loop temperatures also decrease, as shown in Figure 111.5, until the AFW flow is controlled to maintain the specified level in the intact steam generators. These effects result in a decrease in the RCS pressure and pressurizer level, and the pressurizer level goes offscale low. However, the increased SI flow subsequently causes the RCS r,ressure and pressurizer level to increase again. It is assumed that the time required for the operator to identify that 30 minutes. Thus, the isolation of the ruptured steam generator is completed at 2440 seconds and the depressurization of ruptured steam generator is terminated. At this time, the ruptured steam generator pressure increases rapidly 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 and the pressurizer level increases and returns onseale.

2. Cool Down the RCS to establish Subcooling Margin AoL After the a5 minute operator action time is imposed prior to initiation of cooldown. The depressurization of the ruptured steam generater affects 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 cooled by dumping steam to

! the atmosphere using the intact steam generator PORVs. The cooldown is continued until RCS subcooling at the ruptured steam i

11s7v:1o/0219as 30

generator pressure is 20'F plus an allowance of 20*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 2742 seconds and is completed at 3664 seconds.

4 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 dureases during this cooldown process due to shrinkage of the reactor coolant as shown in Figure 111.1.

3. Depressurize to Restore Inventory After the RCS cooldown, a 4 minute operator action time is included prior to depressurization. The RCS is depressurized at 3904 seconds to assure adequate coolant inventory prior to terminating SI flow. With the RCPs stopped, normal pressuri:er spray is not available and thus the RCS is depressurized by opening a pressurizer PORV. The depressurization is continued l until any of the following conditions are satisfi',d: RCS pressure is less than the ruptured steam generator pressure and pressurizer level is greater than the allowance of 4% for pressurizer level uncertainty, or pressurizer level is greater than 77%, or RCS subcooling is less than the 20*F allowance for subcooling uncertainty. The RCS depressurization reduces the break flow as shown in Figure !!!.7, and increases SI flow tc, 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.

11s7v:1o/021988 31

t 1

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 20'F allowance for j 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 4%

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

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

After depressurization is completed, an operator action time of 1 minute 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 I

in Figure III.1. The differential pressure between the RCS and the ruptured steam generator is shown in Figure !!!.6. Figure ,

III.7 shows that the primary to secondary leakage continues after l the SI flow is stopped until the RCS and ruptured steam generator pressures equalize.

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

For this case, the water volume in the ruptured steam generator is 3

significantly less than the total steam generator volume of 5759 f t f 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 III.9.

l .

I b

• l

$ i i  !

i k

n i m o m i..: 32

4 s 4

TABLE III.1 DCPP SGTR ANALYSIS SEQUENCE OF EVENTS .

OFFSITE RADIATION DOSE ANALYSIS EVENT--

TIME (sec)

SG Tube Rupture 0 Reactor Trip 116 SI Actuated 174 Ruptured SG !solated 636

- - 4,4 638

- - 4.4 2440 E

RCS Cooldown Initiated 2742 RCS Cooldown Terminated 3664 RCS Depressurization Initiated 3904 i

RCS Depressurization Tc.rminated 4034 l SI Terminated 4094 ,

i t Break Flow Terminated 4718 [

P i .

l; 4

i I

l i 3  !

1157v;1o/021964 33

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE A0ALYSIS -

i RCS PRESSURE 2400.-

f 22C3.

2000.

G Q.

~ $0E0.

N G

d .

g 1603..

8 u

1400,-

1203.' ,

1C00. 2000. 5223. 4200. SC00.

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T!NC 15CCI t

i a

Figure !!!.1 RCS Pressure - Offsite Radiation Dose Analysis (d

• I i

itshao/otissa 34 i

?

l r

L O!ABLO CANYON STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS SECONDARY PR.TSSURE 1400.

. 1200.'

23rTAcT sG h1000.'

E 920.

e ,

W ,

S I 620,' RUPMtID sc i

d 420.' i i

te2..

'O . 1000. 2000. 5000. 4000. 5000, i TIME ISCCI Figure !!!.2 Secondary Pressure - Offsite Radiation Dose Analysis ,

i i

I r

iistr.io41tsaa 35 n

e C

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE i

0FFSITE DOSE ANALYSIS PRESSURIZER LEVEL 78.' i i

68.' I

\

50.'

5

.J W 40.

3 E '

g53..

m 2e..

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l TINC ISCC)

Figure !!!.3 Pressurizer Level - Offsite Radiation Dese Analysis [

t i

i us7<.tivensas 36

4 i

DIABLO CANYON STEAN GENERATOR TUBE RUPTURE OFFSITE DOSE ANA'LYS!$

RUPTURE 0 LOOP HOT AND COLO LEO RCS TEMPERATURES <

f 650.

t

/

603. "

.z o

550.' Thet c 503.-

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4C2.- ,

u I -

seg*,

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+

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• C . 1220. 22:2. 5000. 4000. 5003. l T!NC ISCCI l

F Figure !!!.4 Ruptured Loop Hot and Cold Leg RCS Teeperatures - l Offsite Radiation Dose Analysis l l

l t

I L

usreinanssa 37 I

k DIABLO CANYON STEAN GENCRATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS INTACT LOOP HOT AND COLO LEO RCS TEMPERATURES 458.<

s 600.<

Thot

$552.-

Tcoid

- 500.

8 5

[450.' Thot

". Teold h400. [

l 2 553 <

502.O. 1200. 2223. 5003. 4000. 5000. l TIME ISCCI L

Figure III.5 Intact Loop Hot and Cold Leg RCS Toeperatures - l Offsite Radiation Dose Analysis  !

I i

L t

11s7 ;1o/ott us 38 I

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS DIFFERENTIAL PRESSURE BETWEEN RCS AND RUPTURED 50 1600.-

1420.

\

1200.- .

1003.-

600.'

d v

622.

t 4:3..

2C3.'

C. l 2C3 2. 1C:3. 2C D. EC C. 4 ::. 5::3.

TIMC ISEC e p

^

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

i iishan/oitsu 39 (

l y

DIABLO CANYON STEAM GENERATOR TUSE RUPTURE ,

OFFSITE DOSE ANAL.'/ SIS PRIMARY TO SECONDARY BMAK FLOW 40.<

60.<\

73.- .

62.<

u d

50.-

'5.

g 40.-

w '

d 32.

W a

20. .

10. 4

0. -

I

'C . 1000. 200J. 5000. 4000. 3 03.

TIME (SEcl [

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

Offsite Radiation Dose Analysis ,

t I

l t

40 l iis7r.io/otissa l

i l

l I

1 1

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE j OFFSITE DOSE ANALYSIS RUPTURED SG WATER VOLUME - ,

5500.-

5220.-

s

{4503.-

4003.'

s '

M 5523.-

3 <

I 8

c !C23.

=

h2523.' l 22:3.- -

isa .. im. m e. im. deco. Sm.  ;

TIME ISECi .

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

I, us7,;toa nsee 41 f

e I

O!A8LO CANYON STEAM GENERATOR TUBE RUPTURE .

I 0FFSITE DOSE ANALYSIS t

, RUPTURED SG WATER MASS 26 228.'

26C;;3.<

24C002.'

5

- 222003.

m 2C2:C2.

1 5Is::::.

8 I o 16:003.-

b 14 003.-

1222C2.4 10:003.<

OC228 '8. 12 3. 2200. 5000. 4CC3. SODO. f TIMC ISCCI {

L Figure !!!.9 Ruptured SG Water Wass - Offsite Radiation  !

Dose Analysis  ;

i I

l f

f im.mune 42 f

f

, 1 f

, 5. Mass Releases l 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, r.nd primary to l secondary break flow into the ruptured s' team generator were determined i 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 /> afi;er the accident. The releases for [

i 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 i for the duration of the accident. ,

i In the LOFTTR2 analyses, the SGTR recovery actions in Diablo Canyon Emergency Procedure E-3 were simulated until the termination of I primary to secondary leakage. After the primary to secondary leakage is terminated, the operators will continue tha SGTR recovery actions to prepare the plant for cooldown to cold shutdown conditions. When these recovery actions are completed, the plant should be cooled and .

depressurizedtocoldshutdoowconditions. - ,

it was assumed tnat tne cooldown is  :

performedusingDiabloCan7enEmergencyProcedureES-3.3, POST-SG1R C00LOOWN USING STEAM DUMP, since this method results in a conservative ,,

evaluation of the long tarfa mass releases for the offsite dose analysis. '

i The high level actions for the the post-SGTR cooldown method using  ;

i steam dump in Diablo Canyon Emergency Procedure ES-3.3 are discussed l I below.  !

1. Prepare for Cooldown to Cold Shutdown [

f The initial steps to prepare for coolocan to cold shutdoan will be i continued if they have not already been ecmpleted. A few additional steps are also performed prict to initiating cooldown. j l

11s7v.1D/c2194: 43 4

These include isolating the cold leg SI-accumulators to prevent unnecessary injection, energizing pressurizer heaters as necessary j to saturate the pressurizer water'and to provide for better  ;

pressure control, and assuring adequate shutdown margin in the l 4 event of potential boron dilution due to in-leakage from the  !

ruptured steam generator. ,

(

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

generators similar to a normal cooldown. Since all immediate j sr.foty concernt have been resolved, the cooldown rate should be i maintained less than the maximum allowable rate of 100'F/hr. The [

p-oferred 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 [

i also be used if steam dump to the condenser is unavailable. Since i i

a loss of offsite power is assumed for the analysis, it was l

! assumed that the cooldown is performed using steam dump to the [

i atmosphere via the intact steam generator PORVs. When the RHR system operating temperature is reached, the cooldown is stepped (

until RCS pressure can also be decreased. This ensures that the }

! pressure / temperature limits will not be exceeded.  !

' [

I 3. Depressurize RCS to RHR System Pressure

.! j When the cooldown to RHR system temperature is completed, the

pressure in the ruptured steam generator is decreased by releasing  !

! steam from the ruptured steam generator. Steam release to the f condenser is preferred since this minimizes radiological releases, j but steam can be released to the atmosphere using the PORY en the l l ruptured steam generator if the condenser is not available, j 1 Consistent with the assumption of a loss of offsite power, it was t

! assumed that the ruptured steam generator is depressurized by l

! releasing steam via the PORV. As the ruptured steam generator f i I m ,o o,m saa u j i

,__.--_-J

4 pressure is reduced, the RCS pressure is maintained equal to the pressure in the ruptured steam ger. orator,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 PORV or auxiliary spray can be used to control RCS pressure if pressurizer spray is not available.

4. Cool Down 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 Diablo Canyon analysis. 4 e methodology and the results of the calculations are discussed balow.

a. Methodology for Calculation of Mass 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 steam generators, and the primary tu 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.

Following the termination of leakage, it was assumed that the RCS andintactsteamgneratorconditionsaremaintainedstablefora unhilthecooldownisinitiated. The PORVs for the intact steam generators were then assumed to be used to cool 11s7v:1o/02198 45

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 temperatures 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

.g, steam releases and the fendwater flows for the intact steam generator for the period from leakag'e termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> weredeterminedfrom{

$.1nce the rupturec steam generatoris'solated,nochangeintEerupturedsteamgenerator conditions is assumed to occur until subsequent depressurization.

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

Depressurization of the ruptured steam generator was then assumed to be performed immediately following the completion of the RCS cooldown. The ruptured steam generator was assumed to be depressurized to the RHR in-service pressure of 405 psia via steam

release from the ruptured steam generator PORV, since this l

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 i is assumed that the continuation of the RCS cooldown and I 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 I 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 l

_ a c.

thesteamreleasedfromtherupturedsteam generator from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was determined based on._

~AC s 115/v:1o/021988 46 l

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

b. Mass Release Results The mass release calculations were performed using the methodology discussed above. For 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 ejector. After reactor trip, the releases to the atmosphere are assumed to be via the steam generator PORVs. The mass release rates to the atmosphoro from the LOFTTR2 analysis are presented in Figures III.10 and III.11 for the ruptured and intact steam generators, respect' N , for the time period until leakage termination.

The mast, releases calculated from the time of leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and from 2-8 hours are also assumed to be released to the atmosphsre via the steam generator PORVs. The mass releases for the SGTR event for each of the time intervals considered are presented in Table III.2. The mass releases prior to break flow termination, from break flow termination until 2 ,

hours, 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 summari7ed in Table !!I.3. The results indicate that approximately 146,700 lbm 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 291,800 lbm of primary water is v ,

. transferred to the secondary side of the ruptured steam generator before the break flow is terminated.

I 11s7tlo/021988 47

TABLE 111.2 DCPP SGTR ANALYSIS MASS RELEASES E OFFSITE RADIATION DOSE ANALYSIS f

TOTAL MASS FLOW (POUNDS)

! TIME PERIOD 0-TRIP TRIP - TMSEP - TTBRK - T2 HRS -

TMSEP TTBRK T2 HRS TRHR Ruptured SG g Condenser 127,100 0 0 0 0 m Atmosphere 0 145,100 1600 0 37,800 Feedwater 117,600 32,000 0 0 0 j! Intact SGs Condensor 375,500 0 0 0 0 76,400 215,300 901,400 li -

Atmosphere 0 153,100 Feedwater 375,500 333,400 132,800 229,200 906,200 l5 -

!(

lj Break Flow 9500 228,700 53,600 0 0 h-

] TRIP = Time of reactor trip = 116 sec.

I TMSEP = Time when water reaches the mo'<sture separators = 3230 sec.

TTBRK = Time when break flow is terminated = 4718 sec.

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

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

46 1157v:1o/021988

=

TABLE III.3 OCPP SGTR ANALYSIS SUMMARIZED MASS RELEASES OFFSITE RADIATION DOSE ANALYSIS TOTAL MASS FLOW (POUNDS) 0- TTBRK - 2HR$ -

TTBRK 2 HRS 8 HRS Ruptured SG Condenser 127,100 0 0 Atmosphere 146,700 0 37,800 Feedwater 149,600 0 0 Intact SGs Condenser 375,700 0 0 Atmosphere 229,500 215,300 901,400 Feedwater 841,700 229,200 906,200 Break Flow 291,800 0 0 1157v:10/021988 49

l DIABLO CANYON STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE ANALYSIS RUPTURED SG ATt10 SPHERIC f1 ASS RELE ASES 225.

U 200.

L' 5

]175.

E 150.

s 2 125.

M

=

y 100.

S E

E 75.

e 50.

!g 25. \

~

O .0. 1000. 2000. 5000. 4000. 5000.

TIME ISEC)

Figure 111.10 Ruptured SG Wass Release Rate to the Atmosphere -

Offsite Radiation Dose Analysis 1157v:10/01158: 50 I

l ,,

s CIABLO CANYON STEAM GENERATOR TUBE RUPTURE OFFSITE DOSE. ANALYSIS INTACT SOS ATMOSPHERIC MASS RELEASE (LB/SEC) 700.<

l I

G (600.

5 N502.

b u

j y400.

t l u

] E l

Y 500.

S ,

l E

2 e 200.

8

, 0 5:00.- (

5 .

I l

20. 1000. 2000. 5000. 4000. 5000.

TIFE (SEC)

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

Offsite Radiation Dose Analysis 1157v:1D/011544 51

i B. 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 radionuclides in the react'or 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 air ejector exhaust.

The quantity of radioactivity released to the environment, due to an SGTR, depends upon primary and secondary coolant activity, iodine spiking effects, primary to secondary break flow, break flow flashing fractions, attenuation of iodine carried by the flashed portion of the break flow, partitioning of iodine between the liquid and steam phases, the mass of fluid released from the generator and liquid-vapor partitioning in the turbine condenser hot well. All of tnese parameters were conservatively evaluated in a manner consistent with the recommendations of Standard Review Plan Section 15.6.3.

1. Design Basis Analytical Assumotions The major assumptions and parameters used in the analysis are itemized in Table III.4.

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.

1157v:1o/021988 52

i. Accident Initiated Spike - The initial p-imary coolant iodine concentration is 1 pCi/gm of Dose Equivalent-(0.E.) I-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 initial appearance rate can be written as follows:

P$=Ag A g

where:

P$ = equilibrium appearance rate for iodine nuclide i A$ = equilibrium RCS inventory of iodine nuclide i corresponding to 1 uCi/gm of 0.E. I-131 kg = removal coefficient for iodine nuclide i

- 4,6 The duration of the spike, is sufficient to ,,, ,,, a, e increase the initial RCS I-131 inventory by a factor of ii. 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 0.E. I-131.

b. The initial secondary coolant iodine concentration is 0.1 uCi/ gram of 0.E. 1-131.

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

1157v:1o/022388 53

d. The initial noble gas concentrations in the reactor coolant are based upon 1% fuel defects. The noble gas inventories, upon which the concentrations are based, were taken from Table 11.1-1 of the Diablo Canyon FSAR.

. 3. Dose Calculations _

The iodine transport model etilized in this analysis was proposed by Postma 46d Tam (Reference 3). The model considers break flow flashing, droplet size, bubble scrubbing, steaming, and partitioning.

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

Removal credit is taken for scrubbing of iodine contained in the atomized coolant droplets when the rupture site is below the secondary water level. The fraction of primary coolant iodine which is not assumed to become airborne immediately 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 conservatively assumes an iodine partition coefficient of 100 between the steam generator liquid and steam phases when the rupture site is covered. The model takes no scrubbing or mixing credit when the rupture site is above the secondary water level. Oroplet removal by the dryers is conservatively assumed to be negligible. The iodine transport model is illustrated in Figure 111.12.

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

4

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

11s7v:1o/0223sa 54

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 rupture site to the water surface conservatively assumes that the rupture is located at- the insersection 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 relative to the top of the tubes in the ruptured and intact steam generators is shown in Figure 111.14. As noted from Figure 111.14, the water level in both the ruptured and intact steam generators drops below the top of the tubes after reactor trip, but then begins to increase and recovers the top of the tubes a short time later. The iodine scrubbing efficiency is determined by the method suggested by Postma and Tam j (Ref. 3). The iodine scrubbing efficiencies are shown in Figure III.15.

l The activity released to the environment by the flashed rupture 1

flow can be written as follows:

1 A =

r IA3 (1 - eff3) j where:

A r

= total iodine released to the environment by flashed primary coolant

=

(integrated activity in rupture flow during time IA) interval j) (flashing fraction for time interval j) 1157v:10/0223:s 55

eff j = iodine scrubbing efficiency during time interval j

d. During the time period that the rupture (or leakage) site is uncovered, all of the activity carried by the break (leakage) flow is assumed to be directly released to the environment, i.e., the activity carried by the break (leakage) flow will neither mix with the secondary water nor partition. The rupture site is considered to be covered when the secondary water level is approximately 12 inches over the rupture site (approximately 8 inches over the apex of the tube bundle).

e. The total primary to secondary leak rate is assumed to be 1.0 gpm as allowed by the Technical Specifications. The leak rate is assumed to be 0.70 gpm for the three intact steam generators and 0.3 gpm for the ruptured steam generator. The leakage to the intact steam generators is assumed to persist for the duration of the accident.

f. The iodine partition coefficient between the liquid and steam of the ruptured and intact steam generators is assumed to be 100 during the time that the rupture (or leakage) site is covered,

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 boundary of the low population zone.

h. Short-term atmospheric dispersion factors (x/Qs) for accident analysis and breathing rates are provided in Table III.8. The breathing rates were obtained from NRC Regulatory Guide 1.4, (Ref.

4).

l 1157v;10/oz1ssa 56

4. Offsite Oose Calculation Offsite thyroid doses are calculated using the equation:

O = DCF Th $[(IAR)$3 (BR)3 (x/0)3 i j l

where (IAR)g3 = integrated activity of iodine'nuclide i released during the time interval j in Ci*

(BR). = breathing rate during time interval j in 3

meter /second (Table III.8)

(x/0)j = atmospheric dispersion factor during time interval j in seconds / meter 3 (Table III.8)

(DCF)$

= thyroid dose conversion factor via inhalation for iodine nuclide i in rem /Ci (Table III.9)

O Th

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

Og

=0.25[ f ri j

(IAR)q) (x/Q)3 i

• Nocreditistakenforclouhdepletionbygrounddepositionorby radioactive decay durint transport to the exclusion area boundary or to the outer boundary of the low population zone.

1157v;1o/021988 57 e

where:

= integrated activity of noble gas nuclide i (IAR)33 released during time interval j in Ci *

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

interval j in seco'nds/m 3 2 ,= average gama energy for noble gas nuclide i in 73 Mov/ dis (Table III.10)

O r = whole body gama dose due to immersion in rem Offsite beta-skin doses are calculated using the equation:

0 3 =0.23{ 2 33 j

(IAR)33 (x/0) i-where:

(IAR)gj

= integrated activity of noble gas nuclide i released during time interval j in Ci *

= atmospheric dispersion factor during time (x/0)) interval j in seconds /m 3

2 = average beta energy for noble gas nuclide i in 33 Mev/ dis (Table III.10)

D

= beta-skin dose due to imersion in rem 3

• 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.

1157v.1o/c21sta 58

A

5. Results Thyroid, whole-body gama, and beta-skin doses at the Exc.lusion Area Boundary and the outer boundary of the Low Population Zone are presented in Table 111.11. All doses are within the allowable guidelines as specified by Standard Review Plan 15.6.3 and 10CFR100.

l l

l J

1157v:10/021988 59

TABLE III.4 DCPP SGTR ANALYSIS PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSE0VENCES i

I. Source Data A. Core power level, MWt 3568 B. Total steam generator tube 1.0 leakage, prior to accident, gpm C. Reactor coolant activity:

1. Accident Initiated Spike The initial RC iodine activities based on 1 uCi/ gram of 0.E. 1-131 are presented in Table 111.5. The iodine appearance rates assumed for the accident initiated spike are presented in Table III.6.

2. Preaccident Spike Primary coolant iodine activities based on 60 uCi/ gram of D.E. 1-131 are presented in Table 111.5.

s 3. Noble Gas Activi.ty The initial RC noble gas activities based on 1%

fuel defects are presented in Table III.7.

11s7v.1o/02198s 60 .

\

TABLE III.4 (Sheet 2)

D. Secondary system initial activity Dose equivalent of 0.1 uCi/gm of I-131, presented in Table 111.5.

l= E. Reactor coolant mass, grams 2.57 x 10 8 l

F. Initial Steam generator mass 4.3 x 10 7 l (each), grams l

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

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

1. Rupture flow See Table III.2

2. Rupture flow flashing fraction See Figure III.13

3. Iodine scrubbing efficiency See Figure III.15

4. Total steam release, lbs See Table III.2 1*

5. Iodine partition coefficient 100 when rupture site is covered 11s7v:1o/022388 61

TABLE III.4 (Sheet 3) '

6. Location of tube rupture Intersection of outer tube row and upper ,

anti-vibration bar

.. B. Intact steam generators -

1

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

2. Total ste4Gi roleese, lbs See Table III.2

3. Iodine partition coefficient 100 when leakage :ite is covered 'N

~'

C. Condenser

1. Iodine partition coefficient 100 D. Atmospheric Dispersion Factors See Table III.8 e

a T 1167v:10/o22388 62

TABLE III.5 DCPP SGTR ANALYSIS IODINE SPECIFIC ACTIVITIES .

IN THE PRIMARY AND SECONDARY COOLANT BASED ON 1, 60 AND 0.1 uCi/ gram 0F D.E.1-131*

Specific Activity (uCi/gm) {

Primary Coolant Secendary Coolant Nuclide 1 uCi/gm 60 uCi/cm 0.1 uCi/gm I-131 0.69 41.4 0.069 I-132 0.25 15.0 0.025 1-133 0.94 56.4 0.094 I-134 0.13 7.8 0.013 1-135 0.52 31.2 0.052 l

l

• Consistent with the DCPP Technical Specificaticns.

2-1157v.10/021988 63

TABLE III.6 DCPP SGTR ANALYSIS IODINE SPIKE APPEARANCE RATES (CURIES /SECOND) s I-131 1-132 1-133 I-134 I_-135 1.54 3.21 3.10 3.82 3.04 1157v;10/021988 64

TABLE III.7 4

DCPP SGTR ANALYSIS NOBLE GAS SPECIFIC ACTIVITIES IN THE REACTOR COOLANT BASED ON 1% FUEL DEFECTS Nuclide Specifi'c Activity (uCi/gm) ,

Xe-133m 3.11 Xe-133 269.5 Xe-135m 0.38 ,'

Xe-135 5.42 Xe-138 0.52 Kr-85m 1.8 Kr-85 5.7/

Kr-87 0.98 Kr-88 3.07 , i l

e 1157v:10/02198: 65

TABLE III.8 DCPP SGTR ANALYSIS ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES Time Exclusion Area Boundary Low Population Breathing 3 3 3 (hours) x/0 (Sec/m ) Zone x/0 (Sec/m ) Rate (m /Sec) [4]

7

~4 -5 -4 0-2 5.29 x 10 2.2 x 10 3.47 x 10

-5 -4 2-8 -

2.2 x 10 3.47 x' 10

/

A4 1157v:10/021958 66 ,

TABLE III.9 -

DCPP SGTR ANALYSIS THYROID DOSE CONVERSION FACTORS (Rem / Curie) (Ref. 5)

Nuclide ,

6 I-131 1.49 x 10 4

1.43 x 10 I-132 5

I-133 2.69 x 10 3

I-134 3.73 x 10 4

1-135 5.60 x 10

. a.

1157v:10/021988 67

TABLE 111.10

~

DCPP SGTR ANALYSIS AVERAGE GAMMA AND BETA ENERGY FOR NOBLE GASES ,

(Mev/ dis) (Ref. 6)

Nuclide h

(

Xe-133m 0.02 0.212 Xe-133 0.03 0.153 Xe-135m 0.43 0.099 Xe-135 0.246 0.325 .'

Xe-138 1.2 0.66 Kr-85m 0.156 0.253 -

Kr-85 0.0023 0.251 Kr-87 0.793 1.33 Kr-88 2.21 0.248 .

s '

i l

l 1157v:10/021988 60

TABLE III.11 OCPP SGTR ANALYSIS  ;

0FFSITE RADIATION DOSES Doses (Rem)

Calculat'ed Allowable ..

Value Guideline Value (Ref. 7]

1. Accident Initiated lodine Spike Exclusion Area Boundary (0-2 hr.)

Thyroid Dose 28.8 30 .

j Low Population Zone (0-8 hr.)  ;.

Thyroid Dese 0.3 30 .

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

Thyroid Dese 192.4 300 Low Population Zone (0-8 hr.)

Thyroid Dese 8.0 300 s

3. Whole-Body Gamma and Beta-Skin Oose Exclusion Area Boundary (0-2 hr.) ,

Whole-Body Gamma Dose 0.23 2.5*

Beta-Skin Dese 0.58 2.5*

e Low Population Zone (0-8 hr ) '

- Whole-Body Gamma Dose 0.01 2.5* .

Beta-Skin Dese 0.02 2.5*

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

11s7v:1o/02238: 69

e OROPWT5 NOT scmuseco Yo" VAPOM &

-9 SCMUSSIN O C PR1M 15 vts S gMcAx - scmusato T u omoes COVEREO? E wsTER A NOT e M A FLASHED -e SECONDARY  ;  ; y WATER 3 y P O ,

. A s e P E H r

. n.nsn n .

INTO E won s DROPATS SPRAY NOT g 3 R EAKU P  :

FLASHED INTO OROPS e rammen Figure III.12 lodine Transport Model - Offsite Radiation Dose Analysis ,

1157r10/021ss: 70

DIABLO CANYON STEAM GENERATOR TUBE RUPTURE OFFSITE 00SE ANALYSIS BREAK FLOW FLASHING FRACTION

.2

/

.19<

.16-g.14-g .12-5 g .i.

$.20-

.CG-ca

.Co

.02

'O . 1003. 2200. 5000. 4000. 5003.

i i TIME ISEC1 Figure III.13 Break Flow Flashing Fraction - Offsite Radiation Dese Analysis 1157v:10/021344 71

DIABLO CANYON STEAN GENERATOR TUBE RUPTURE.

OFFSITE DOSE ANALYSIS SO SECONDARY LEVEL ABOVE TOP OF TUBES 500.-

E

250.

0 8

223' IturmRzo so 9

y 150.

5 8

d i00.. mmu 5

50.-

U U

g 6.

'O . 1000. 2000. 5000. 4000. 5000.

TIME ISEcl Figure 111.14 SG Water Level Above Top of Tubes -

Offsite Radiation Dose Analysis 1157v:1o/021388 72

\

0.05 l DIRELO CANYON 53TR 0.05 -

O.01 -

, W ,

i w o.e2 -

u I E 0.02 -

E

$ 0.01 - v 2

u C ,l

. i I l I i g,g) 0 930 1000 15CO 2000 150] 2000 TIME (55 CON 05)

Figure 111.15 lodine Scrubbing Efficiency - Offsite Radiation Dose Analysis 1157v:10/021984 73

IV. CONCLUSION An evaluatb. has been performed for a design basis SGTR event for the Diablo ,

Canyon 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 tc, overfill.

The limiting single failure is the failure of -

ae theresultsofthis analysisindicatethattherecoveryactionscanbecompletedtoterminatethe primary to secondary break flow before overfill of the ruptured steam generator would occur.

Since.it is concluded that steam generator overfill will not occur for a design basis SGTR, an analysis was also performed to determine the offsite radiation doses assuming the limiting single failure for offsite doses. For this analysis, it was assumed that the

- 4 , d, The primarytosecondarybreakflowandthemassreleascatotheatmosphere7ere 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 Diablo Canyon would be acceptable.

11s7v:1o/0219ss 74

V. REFERENCES i.

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

f l

2. Lewis, Huang, Rubin, "Evaluation of Offsite Radiation Doses for a Steam l

Generator Tube Rupture Accident," Supplement 1 to WCAP-10750-A, March 1986. i l

1

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

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

1

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

6. Bell, M. S. "0RIGEN - The ORNL Isotope Generation and Depletion Code",

ORNL-8628, 1973.

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

tis 7v.to/on ses 75

. _ _ _ _ _ _ _ _ _ _ _