ML17256A402

Analysis of Plant Response During 820125 Steam Generator Tube Failure at Re Ginna Nuclear Power Plant.

ML17256A402
Person / Time
Site:	Ginna
Issue date:	11/22/1982
From:	Volpenhein E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17256A400	List: ML17256A399 ML17256A401 ML17256A402 ML17256A403 ML17256A404 ML17256A405
References
3274Q:1-111782, NUDOCS 8211290429
Download: ML17256A402 (91)

Text

ATTACHMENT A ANALYSIS OF POTENTIAL ENVIRONMENTAL CONSEQUENCES FOLLOWING A STEAM GENERATOR TUBE FAILURE AT R. E. GINNA NUCLEAR POWER PLANT NOVEMBER 1982 Prepared by:

K. Rubin E. Volpenhein Westinghouse Electric Corporation Nuclear Energy Systems P;0. Box 355 Pittsburgh, Pennsylvania 15230 Prepared for:

Rochester Gas and Electric 89 East Avenue Rochester, New York 14649 ggffg+O4PP 821122 PDR ADOCK 05000244 P PDR

TABLE OF CONTENTS Section Page ABSTRACT ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 LIST OF TABLES FIGURES.................... ~...., ..

.....

LIST OF iv I. INTRODUCTION ~ 1 II. MASS RELEASES ~ ~ ~

2 II.l Design Basis Accident . ~ ~ ~ 2 II.l.l Sequence of,Events ~ ~ ~ 2 II.1.2 Method of Analysis ~ ~ ~ 5 II.2 Ginna Event . 10 II I. ENVIRONMENTAL CONSEQUENCES ANALYSIS ~ ~ ~ ~ ~ ~ ~ 27 III.l Design Basis Accident ~ ~ ~ ~ ~ ~ 0 27 III.2 Ginna Event Analysis . ~ ~ ~ ~ ~ ~ o 4D IV.

SUMMARY

AND CONCLUSIONS ~ ~ ~ ~ ~ ~ o 56

'REFERENCES

' 'i ~ ~ ~ ~ ~ ~ ~ ~

57

ABSTRACT The potential radiological consequences of a steam generator tube failure event were evaluated for the R. E. Ginna nuclear power plant to demonstrate that standard limitations on initial coolant activity are acceptable. Mass releases following a design basis tube rupture were calculated for both 30 minute and 60 minute operator response times. The site boundary and low population zone exposures were conservatively calculated for these releases.

'n addition, the standard technical specification limit on initial coolant activity and realistic meteorology were applied to "best estimate" mass "release during the January 25, 1982 tube failure event at Ginna. Results show that the conservative assessment of the environmental consequences are within acceptable limits and that the potential exposure from a more realistic event is minimal.

LIST OF TABLES TABLE II.1.2-1 DESIGN BASIS ACCIDENT SEQUENCE OF EVENTS TABLE II.1.2-2 . MASS RELEASES DURING A DESIGN BASIS SGTR: 30 MINUTE RECOVERY TABLE II.1.2-3 MASS RELEASES DURING A DfSIGN BASIS SGTR: 60 MINUTf RECOVERY TABLE II.2-1 GINNA SEQUENCE OF EVENTS TABLE II.2-2 BEST ESTIMATE MASS RELEASES DURING GINNA SGTR EVENT TABLE III.1-1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE TABLE III.1-2 IODINE APPEARANCf RATES IN THE REACTOR COOLANT FOR A,DESIGN BASIS SGTR TABLE III.1-3 REACTOR COOLANT IODINE AND NOBLE GAS ACTIVITY TABLE III.1-4 SHORT-TERN ATMOSPHERE DISPERSION FACTORS AND BREATHING RATES FOR ACCIDENT ANALYSIS TABLE I II.1-5 ISOTOP IC DATA TABLE III.1-6 RESULTS OF DESIGN BASIS ANALYSIS TABLf III.2-1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF THf GINNA EVENT TABLE III.2-2 IODINE APPEARANCE RATES IN THE REACTOR COOLANT

LIST OF TABLES (Continued)

TABLE III.2-3 SHORT-TERM ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES FOR ACCIDENT ANALYSIS TABLE III.2-4 RESULTS OF GINNA EVENT ANALYSIS 111

LIST OF FIGURES FIGURE II.1.2-1

~ ~ FAULTED STEAM GENERATOR WATER VOLUME FIGURE II.1.2-2 REACTOR COOLANT SYSTEM PRESSURE FIGURE II.1.2-3 FAULTED STEAM GENERATOR PRESSURE FIGURE II.l . 2-4 REACTOR COOLANT AVERAGE TEMPERATURE FIGURE II.1.2-5 PRESSURIZER WATER VOLUME FIGURE II.1.2-6 FAULTED STEAM GENERATOR STEAM FLOW FIGURE I I.l. 2-7 PRIMARY-TO-SECONDARY LEAKAGE FIGURE II.1.2-8 BREAK FLOW FLASHING FRACTION FIGURE II.2-1 CALCULATED FAULTED STEAM GENERATOR WATER VOLUME DURING THE GINNA EVENT FIGURE I I. 2-2 REACTOR COOLANT SYSTEM PRESSURE DURING THE GINNA EVENT FIGURE II.2-3 FAULTED STEAM GENERATOR PRESSURE DURING THE GINNA EVENT FIGURE II.2-4 CALCULATED BREAK FLOW FLASHING FRACTION DURING THE GINNA EVENT FIGURE III.l-l BREAK FLOW FLASHING FRACTION FOR THE DESIGN BASIS EVENT DOSE ANALYSIS FIGURE III.1-2 'TTENUATION FACTOR FOR FLASHED COOLANT FOR THE DESIGN BASIS EVENT DOSE ANALYSIS

'IST OF FIGURES (Continued)

FIGURE III.1-3 FAULTED STEAM GENERATOR PARTITION FACTOR FOR THE DESIGN BASIS EVENT DOSE ANALYSIS FIGURE III.2-1 BREAK FLOW FLASHING FRACTION FOR THE GINNA EVENT DOSE ANALYS IS FIGURE III.2-2 ATTENUATION FACTOR FOR FLASHED.COOLANT FOR THE GINNA EVENT DOSE ANALYSIS FIGURE III.2-3 FAULTED STEAM GENERATOR PARTITION FACTOR FOR THE GINNA EVENT DOSE ANALYSIS

I. INTRODUCTION Potential environmental consequences of a steam generator tube rupture event at the R. E. Ginna nuclear power plant have been evaluated to verify. that the standard technical specification limit on primary coolant activity is ade uate for Ginna. Mass releases were calculated using the computer code LOFTRAN with conservative assumptions of break size, condenser availability, and various operator response times. The effect of steam generator overfill and subsequent water relief through secondary side relief valves was also addressed. Conservative assumptions concerning coolant activity, meteorology, and partitioning between liquid and vapor phases were applied to these mass releases to determine an upper bound on site boundary and low population zone doses. Best estimate mass releases during the January 25, 1982 tube failure event at Ginna,were also calculated based on analyses presented in reference

2. These releases were used to estimate potential doses which could have if resulted, the accident had.occurred with coolant activity limits established in the 'standard technical specifications.

II. MASS RELEASES

'ass releases during a design basis steam generator tube rupture event were calculated using established fSAR methodology assuming various operator response times. Releases during the Ginna event were also estimated.

Contributions from both the intact and faulted steam generators were evaluated as well as flow to the condenser and atmosphere. These mass releases are presented for various time periods during the accident. The assumptions and methodology which were used to generate the results +re described in the following sections.

II.l Design Basis Accident The accident examined is the complete severance of a single steam generator tube during full power operation. This is considered a condition IV event, a limiting fault, and leads to an increase in the contamination of the secondary system due to leakage of radioactive coolant from the RCS. Discharge of acti-vity to the atmosphere may occur via the steam generator safety and/or power operated relief valves. The concentration of contaminants in the primary system is continuously controlled to limit such releases.

II.1.1 Sequence of Events If normal operation of the various plant control systems is assumed, the fol-lowing sequence of events is initiated by a tube rupture:

A. The steam generator blowdown liquid monitor and/or the condenser air ejector radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system.

B. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side steam flow/feedwater flow mismatch occurs as feedwater flow to the affected steam generator is reduced to compensate for break flow to that unit.

C. The decrease in RCS pressure due to continued loss of reactor coolant inventory leads to a reactor trip signal on low pressurizer pressure or overtemperature delta-T. Plant cooldown following reactor trip leads to a rapid decrease, in pressurizer level and a safety injection signal, initi-

"ated by low pressurizer pressure, follows soon after reactor trip. The safety injection signal automatically terminates normal feedwater supply and initiates auxiliary feedwater addition.

D. The reactor trip automatically trips the turbine and, if offsite power is available, the steam dump valves open permitting steam dump to the conden-ser. In the event of coincident station blackout, as assumed in the results presented, the steam dump valves automatically close to protect the condenser. The steam generator pressure rapidly increases resulting in steam discharge to the atmosphere through the steam generator safety and/or power operated relief valves.

E. The auxiliary feedwater and borated safety injection flow provide a heat sink which absorbs decay heat and attenuates steaming from the steam gene-rators.

F. Safety injection flow results in increasing pressurizer water volume at a rate dependent upon the amount of auxiliary equipment operating. RCS pressure eventually equilibrates at a pressure greater than the affected steam generator pressure where safety injection flow matches break flow.

The operator is expected to determine that a steam generator tube rupture has occurred and to identify and isolate the faulty steam generator on a restric-ted time scale in'order to minimize contamination of the secondary system and ensure termination of radioactive release .to the atmosphere from the faulty unit. Sufficient indications and controls are provided to enable the operator .

to complete recovery procedures from within the control room. High radiation indications or rapidly increasing water level in any steam generator provide symptoms of the faulted steam generator which ensure identification before the water level increases above the narrow range. For smaller tube failures,

sampling of the steam generators for high radiation may be required for positive identification. However, in that case additional time would be available before water level increases out of narrow range.

Once identified, the faulted steam generator is isolated from the intact steam generators to minimize activity releases and as a necessary step toward estab-lishing a pressure differential between the intact and faulted steam genera-tors. The Mai'n Steamline Isolation Valves (NSIV) provide this capability. In the event of a failure of the MISV for the faulted steam generator, the NSIV for the intact steam generator and the turbine stop valve ensure a redundant means of isolation. Auxiliary feedwater flow is terminated to the faulted unit in an attempt to control steam generator inventory.

The reactor coolant temperature is reduced to establish a minimum of 50 F subcooling margin at the ruptured steam generator pressure by dumping steam from the intact steam generator. This assures that the primary system will remain subcooled following depressurization to the faulted steam generator pressure in subsequent steps. If the condenser is available, the normal steam dump system is used for this cooldown. Isolation of the faulted steam genera-tor ensures that pressure in that unit will not decrease significantly. If the condenser is unavailable or if the MSIV for the faulted steam generator fails, the atmospheric relief valve on the intact steam generator provides an alternative means of cooling the reactor coolant system.

The primary pressure is reduced to a value equal to the faulted steam genera-tor pressure using normal pressurizer spray. This action restores pressurizer level as safety injection flow in excess of break flow replaces condensed steam in the pressurizer, and momentarily stops primary-to-secondary leakage.

If normal spray is not available, the pressurizer PORVs and auxiliary spray system provide redundant means of depressurizing the reactor coolant system.

l Termination of safety injection flow is required to ensure that break flow is not reinitiated. Previous operator actions are designed to establish suffi-cient indications of adequate primary coolant inventory and heat removal so that core cooling will not be compromised as a result of SI termination.

This sequence of recovery actions ensures early termination of primary-to-secondary leakage with or without offsite power available. The time required to complete these actions are event specific since smaller breaks may be more difficult to detect. In these analyses, operator action times have been treated parametrically, ranging from 30 minutes to a maximum of 60 minutes to complete the key recovery sequence.

II.1.2 Method of Analysis Mass and energy balance calculations were performed using LOFTRAN to determine primary-to-secondary mass leakage and the amount of steam vented from'each of the steam generators prior to terminating safety injection. In estimating the mass releases during recovery, the following assumptions were made:

A. Reactor trip occurs automatically as a result of low pressurizer pressure or overtemperature delta-T. Loss of offsite power occurs at reactor trip.

B. Following the initiation of the safety injection signal, all safety injec-tion pumps are actuated. Flow from the normal charging pumps is not con-sidered since it is automatically terminated on a safety injection signal.

C. The secondary side pressure is assumed to be controlled at the safety valve pressure following reactor trip. This is consistent with loss of offsite power.

D. Auxiliary feedwater flow is assumed throttled to match steam flow in all steam generators to control steam generator level. Minimum auxiliary feedwater capacity is assumed. This results in increased steaming from the steam generators.

E. Individual operator actions are not explicitly modeled in the analyses presented. However, it is assumed that the operator completes the recovery sequence on a restricted time scale. This time is treated para-metrically.

F. For cases where steam generator overfill occurs, water relief from the faulted steam generator to the atmosphere is assumed equal to any addi-tional primary-to-secondary leakage after overfill occurs. Steamline volume is not considered in calculating the time of steam generator over-fil 1 .

Prior to reactor trip steam is assumed to be released to the condenser from the faulted and intact steam generators. Steam from all steam generators is dumped to the atmosphere after reactor trip since the condenser is unavailable as a result of station blackout.

Extended steam release calculations, i.e. after break flow has been termina-ted, reflect expected operator actions as described in the Mestinghouse Owners Group's Emergency Response Guidelines . Following isolation of the faulted steam generator, it is assumed that steam is dumped from the intact steam generator to reduce the RCS temperature to 50'F below no-load Tavg.

From two to eight hours after tube failure, the RCS coolant temperature is reduced to Residual Heat Removal System (RHRS) operating conditions via addi-

. tional steaming from the intact steam generator. Further plant cooldown to cold shutdown, is completed with the RHRS. If steam generator overfill does not occur, the faulted steam generator is depressurized by releasing steam from that steam generator to the atmosphere. An alternate cooldown method, such as backfill into the RCS, is considered if the faulted steam generator fills with water. In that case additional steaming occurs from the intact steam generator. The extended steam and feedwater flows are determined from a mass and energy balance including decay heat, metal heat, energy from one operating reactor coolant pump, and sensible energy of the fluid in the RCS and steam generators.

The sequence of events for the design basis accident are presented in Table II.1. 2-1. The primary-to-secondary car ryover and steam and feedwater flows associated with each of the steam generators are provided in Tables II.1.2-2 and II.1.2-3 for recovery times of 30 and 60 minutes, respectively. Since individual operator actions were not modelled, the system response is the same for both cases. Mith 30 minute operator action to terminate break flow,

TABLE II.1.2-1 DESIGN BASIS ACCIDENT SEQUENCE OF EVENTS Manual (0) Time (Sec)

Event Automatic (A) 30 Min Recovery 60 Min Recovery Tube Failure Reactor Trip 27 27 Condenser Lost 27 27 SI Signal 127 127 Feedwater Isolation 134 134 AFW Initiation 187 187 AFW Throttled to Faul.ted SG 187 187 Isolation of Faulted SG 1800(1) 36oo(1)

.

Steam Dump lsoo(1) 3600(1)

RCS Depressurization 1800(1) 3600(l)

SG Overfill 2S10 SI Terminated 1SOO<<) 3600(1)

Break Flow Terminated 1800(1) 3600(1)

RHR Cooling 28800 28800

\

(1) These events are not actually modeled but are assumed to occur within the time indicated.

TABLE II.1.2-2 MASS RELEASES DURING A DESIGN BASIS SGTR: 30 MINUTE RECOVERY Time Period Flow (ibm) 0-TTRIP TTRIP-TTBRK TTBRK-2 2-TRHR Ruptured SG:

- Condenser 27820 0.0 0.0 0.0

- Atmosphere 0.0 32640 0.0 21 480

- Feedwater 32605 0.0 0.0 21480 Intact SG:

- Condenser 27380 0.0 0.0 0.0

- Atmosphere 0.0 23050 144650 470000

- Feedwater 37170 13370 206200 487600 Break Flow 3325 100648 0.0 0.0 TTRIP = 27.0 sec = Time of reactor trip TTBRK = 1800, sec = Time to terminate break flow TRHR = 28800 sec = Time to establish RHR cooling

I TABLE II.1.2-3 MASS RELEASES DURING A DESIGN BASIS SGTR: 60 MINUTE RECOVERY Time Period Flow (ibm) 0-TTRIP TTRIP- TMSEP- TSGOF- TTBRK-2 2-TRHR TMSEP TSGOF TTBRK Ruptured SG:

- Condenser 27820 0.0 0.0 0.0 0.0 0.0

- Atmosphere 0.0 33570 4830 431 71 0. 0 0.0

- Feedwater 32605 0.0 0.0 0.0 0.0 0.0 Intact SG:

- Condenser 27380 0.0 0.0 0.0 0.0 0.0

- Atmosphere 0.0 23370 1390 390 67970 501100

- Feedwater 37170 13700 1390 380 129600 518700 Break Flow 3325 107742 48070 431 71 0. 0 0.0 TTRIP = 27.0 sec = Time of reactor trip TMSEP = 1930 sec = Time to fill SG to moisture separators TSGOF = 2810 sec = Time to fill SG (w/o steamline volume)

TTBRK = 3600 sec = Time to terminate break flow TRHR = 28800 sec = Time to establish RHR cooling 9

liquid level in faulted steam generator remains below the bottom of the mois-ture separator, Figure II.1.2-1. Hence, for this case, partitioning between the vapor and liquid phases effectively reduces radiological releases for the duration of the accident. For delayed recovery, case 2, the moisture separa-tor begins to flood at 32 minutes. The faulted steam generator is completely filled by 47 minutes. During this time, liquid entrainment within the steam flow would increase so that the effectiveness of partitioning would be reduced. Beyond 47 minutes, i.e. steam generator overfill, water relief from the faulted steam generator is assumed equal to break flow.

The following is a list of figures of pertinent time dependent parameters:

FIGURE II.1.2-1 FAULTED SG WATER VOLUME FIGURE II;1.2-2 REACTOR COOLANT SYSTEM PRESSURE FIGURE II.1.2-3 FAULTED SG PRESSURE FIGURE II.1.2-4 REACTOR COOLANT SYSTEM TEMPERATURE FIGURE II.1. 2-5 PRESSURIZER WATER. VOLUME FIGURE II.1.2-6 FAULTED SG STEAM FLOW FIGURE I I.l. 2-7 BREAK FLOW FIGURE II.1.2-8 BREAK FLOW FLASHING FRACTION I I. 2 GINNA EVENT A detailed thermal-hydraulic analysis of the Ginna event is described in reference 2. The results of that analysis form the basis for the calculation of the potential environmental consequences. The general sequence of events during the Ginna accident, Table II.2-1, was similar to the design basis 10

7000.0 6000.0 5000.0 S.G. VOLUf1E

~

I i000.0

~~ 3000.0 I

2000.0 1000.0 0.0 ED CD C) CD CD ED CD C) CD C) CD C7 CD CD Cl CD CD CD EV 40 I TINK (MIN)

FIGURE I I.1.2-1. FAULTED STEAN GE)lERATOR HATER VOLU)1E.

11

2300.0 2250.0 2000.0 1750.0 1500.0

a. 1250.0 1000.0 CL 750.00 500.00 300.00 Cl CD CD CD Cl Cl CD Cl CD Cl Cl Cl CD Cl Cl AJ m I@1 CO TIN'E (MlN)

FIGURE I I. 1 . 2-2. REACTOR COOLAilT SYSTEh PRESSURE.

12

1200.0 1000.0 800.00

- 600 F 00

~ <u0.00 200.00 0.0 CD CD CD CD CD CD CD CD C) CD C7 CD Cl C7 CD CD AJ m I/I CD TIHE (HIM>

FIGURE II.1.2-3. FAULTED STEAH GENERATOR PRESSURE.

13

700.00 500.00 F00.00 Cl

~ 300.00 I

~ 200.00 100.00 0.0 CD CD CD CD CD CD CD CD CD CD CD CD CD CD CD CD C) C4 lPI co T1ME (M1H)

FIGURE I I.l. 2-4. REACTOR COOLANT AVERAGE TEi1PERATURE.

800.00

?00.00 500.00 au F00.00 X

F00.00 100.00 0;0 CI CI CI CI CI CI CI CI CI CI AJ m TlNE (M l N)

FIGURE II.1.2-5. PRESSURIZER HATER VOLUtlE.

15

0.2000 0.1750

0. 1500 O

Q

0. 1250

0. 1000 CD 0.0750 CD 0.0500 0.0250 0.0 CD CD CD CD CI 8 CD CD CD CD CD CD CD m lA CO TIME (HIM)

FIGURE II.1.2-6. FAULTED STEAN GENERATOR STEAN FLOW.

16

150.00 125.00 100.00 l5.000

~ 50.000 25.000 0.0 CD CD CD CD CD 0

CD CD CD CD CD CD AJ m ICl ED TIME (MIN)

FIGURE I I.l. 2-7. PRIl1ARY-TQ-SECONDARY LEANGE.

17

0.2000

0. 1750

0. 1500

0. 1250 I- 0.1000 0.0500 0.0250

'0.0 CD CD CD CD CD CD CD CD CD CD CD CD CD flJ m Vl

\

TIME (MIN)

FIGURE II.1-2-8. BREAK FLOll FLASHING FRACTION.

18

J t TABLE II.2-1 GINNA SE()UENCE OF EVENTS Event Manual (0) Time (sec)

Automatic (A) Actual Simul ated Tube Failure 0 0 Reactor Trip 182 182

'I Condenser Lost Signal Feedwater Isolation A

A 4500 190 192 4500 198 198 ..

AFW Initiated A 220 239 AFW Throttled to Faulted SG 0 410 410 Isolation of Faulted SG 0 890 530 Steam Dump 0 770 530 RCS Depressurization 0 2700 2700 SG Overfill 3130 SI Terminated 0. 4310 4310 Break Flow Terminated 0 10800 10800 RHR Cooling 0 77580 77580 includes steamline volume 19

event described in section II.l.l. Break flow in excess of normal charging

- flow depleted reactor coolant inventory and eventually resulted in reactor trip on low pressurizer pressure. A safety injection signal followed soon after trip. Normal feedwater flow was automatically terminated on the safety injection signal and auxiliary feedwater flow was initiated. The steam dump system operated to control steam gene- rator pressure below the safety valve setpoint and establish no-load reactor coolant temperature. Auxiliary feedwater and'afety injection flows absorbed decay heat and temporarily stopped steam releases from the steam generators.

Emergency recovery actions were quickly initiated to mitigate the consequences of the accident. Pre-trip symptoms of the faulted steam generator, including steam flow/feed flow mismatch and steam generator level deviation alarms, provided tentative indications of the faulted steam generator which were con-firmed soon after reactor trip by rapidly increasing steam generator level and high radiation indications. Auxiliary feedwater flow was reduced to the faulted unit in an attempt to control inventory. Isolation of the faulted steam generator was completed wi thin 15 minutes of tube failure by closing the associated MSIV. Continued auxiliary feedwater flow to the intact steam gene-rator effectively reduced the primary system temperature to establish 50 F subcooling margin. Normal spray was unavailable since reactor coolant pumps were manually tripped soon after reactor trip as directed by emergency proce-dures. Consequently, one pressurizer PORV was used as an alternative means of depressurizing the primary system to restore pressurizer level and reduce break flow. This was completed within 45 minutes. Safety injection flow was subsequently terminated after 72 minutes. Continued charging flow and reini-tiation of safety injection flow resulted in additional primary-to-secondary leakage until approximately 3 hrs after tube failure.

Mass releases during the Ginna event are presented in Table II.2-2. LOFTRAN results indicate that the faulted steam generator and steamline filled with water after approximately 52 minutes, Figure II.2-1. Beyond this time water relief from the faulted steam generator was assumed equal to any additional primary-to-secondary leakage. The measured primary and faulted steam genera-tor pressures and calculated break flow flashing fraction during the accident 20

TABLE II.2-2 BEST ESTIMATE MASS RELEASES DURING GINNA SGTR EVENT Time Period Flow (ibm) 0-TTRIP TTRIP- TMSEP- TSGOF*-2 2-TTBRK TTBRK-TMSE P TSGOF* TRHR Faulted SG:

- Condenser 162100 16900 0 0

- Atmosphere 0 0 130442 105684

- Feedwater 163400 46800 0 , 0 ~

0 Intact SG:

- Condenser 160100 28800 25200 14500 0 0

- Atmosphere . 0 0 . 0 23870 54743 978387

- Feedwater 171700 52300 0 89700 53008 983292 Break Flow 10300 54330 99170 130442 105684 TTRIP = 182.0 sec = Time of reactor trip TMSEP = 1335 sec = Time to fill SG to moisture separator TSGOF = 2192 sec = Time to fill SG TSGOF* = 3131 sec = Time to fill SG and steamline TTBRK = 10200 sec = Time to terminate break flow TRHR = 77580 sec = Time to establish RHR cooling 21

7000.0 6000.0 S.G. AND STEAr>LINE VOLUWE 5000.0 S.G. VOLUtlE F000.0 I

~) 3000.0 I

2000.0 1000.0 0.0 CD CD OO CD CD CD CD CD CD CD CD G)

~

D~

CD IA CD i/I CD CD V1 AJ "

CD tA A (Q O

CD AJ Pea TlHE <HlN)

FIGURE II.2-1. CALCULATED FAULTED STEAH GENERATOR MATER VOLUt1E DURING THE GINNA EVENT.

22

2300.0 2250.0 2000.0 1750.0 1500.0 C

G 1250.0 G G G G

1000.0 .G G

, 0 750.00 500.00 300.00 Cl D CD C) O~ O~

CI Itl ED If) Q AJ IA (o ED Tl ME (Ml N)

FIGURE II 2 2 REACTOR COOLANT SYSTEi~'1 PRESSURE DURIHG THE GIHHA EYEHT.

23

1200.0 cc 1000. 0 800.00

~ 600.00

~ F00.00 CL 200.00 0.0 Cl Cl CD Cl CD OO Cl Cl CD Cl Cl CI CD Cl CD OO~ ~

Cl CD Ill Cl IO O Cl lA Cl I/I Cl AJ ICl AJ TIME (MIN)

FIGURE II.2-3. FAULTED STEAh GENERATOR PRESSURE DURING THE GINNA EVENT.

0.2000 0 i)50 0 0500 0.0250 0 0 CI TENT tNttll FIGURE II.2-4. CALCULATED BREAK FLOli FLASHI(HG FRACTION DURING T)lE GIN(iA EVEiPT.

25

are presented in Figures II. 2-2 thru II. 2-4. These results show that approxi-mately 236,000 ibm of mass were released after the faulted steam generator and steamline was calculated to fill with water. Approximately 130,000 ibm of this were released in the first 2 hrs. Steam flow to condenser was terminated at approximately 75 minutes. Mass releases were terminated when the RHRS was placed in service after 21.5 hrs.

\

26

I I I. ENVIRONMENTAL CONSEQUENCES ANALYSIS In troduc.ti on For the evaluation of the radiological consequences of a steam generator tube rupture, it is assumed that the reactor has been operting with a small percent of defective fuel for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant. Hence, radionuclides from the

'rimary coolant enter the steam generator, via the ruptured tube, and are released to the atmosphere through the steam generator safety or power operated relief valves.

The radioactivity released to the environment, due to a SGTR, depends upon primary and secondary coolant activity, iodine spiking effects, primary to secondary break flow, time dependent break flow flashing fractions, time dependent scrubbing of flashed activity, partitioning of the activity from the non flashed fraction of the bre'ak flow between the steam generator liquid and steam and the mass of fluid discharged to the environment. All of these parameters were conservatively evaluated for a design basis tube failure, i.e.

double ended rupture of a single tube, as described in Section II.1. The mass releases during the Ginna event were also estimated in Section II.2. The environmental consequences at these events were calculated and are discussed in the following sections.

II I.l DESIGN BASES ANALYTICAL ASSUMPTIONS The major assumptions and parameters used in the analysis are itemized in Table I I.l-l and are summarized below.

27

Source Term Calculations The concentrations of nuclides in the primary and secondary system, prior to the accident are determined as follows:

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

i. Preaccident Spike - A reactor transient has occured prior to the SGTR and has raised the primary coolant iodine concentration to 60 pCi/gram of Dose Equivalent I-131.

ii. Accident Initiated Spike - The reactor trip or primary system depressurization associated with the SGTR creates an iodine spike in the primary system which increases the iodine release rate from the fuel to the primary coolant to a value 500 times greater than the release rate corresponding to the maximum equilibrium primary system iodine concentration of lpCi/gram of Dose Equivalent (D.E.) I-131.

The duration of the spike is assumed to be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Iodine appearance rates in the reactor coolant are presented in Table III.1-2. Doses are calculated for both cases of spiking.

b. The noble gas activity in the reactor coolant is based on 1 percent fuel defects, as provided in Table III.1-3.

The assumption of 1 percent fuel defects for the calculation of noble gas activity, is conservative, since lpCi/gram D.E. I-131 and 1 percent defects cannot exist simultaneously. Iodine activity based on 1 percent defects would be greater than twice the Standard Technical Specification limit.

c. The secondary coolant activity is based on the O.E. of 0.1 pCi/gram of I -131.

d. Iodine at the rupture point is assumed to consist of 99.9 percent elemental and 0.1 percent organic iodine.

28

'I Dose Calculations The following assumptions and parameters are used to calculate the activity released and the offsite doses following a SGTR.

a. The mass of reactor coolant discharged into the secondary system through the rupture and the mass of steam and/or water released from the intact and faulted steam generators, to the environment is presented in Tables II.1.2-2 and 3.

b. The time dependent fraction of rupture'flow that flashes to steam and is immediately released to the environment is shown in Figure III-l-l.

c. The time dependent elemental iodine attenuation factor for retention of atomized primary droplets by the moisture separators and dryers and for scrubbing of steam bubbles as they rise from the leak site to the water surface is presented in Figure III.1-2.

Retention by moisture separators and scrubbing are effected by differential pressure (aP) across the ruptured tube and water level., Specifically for the first 4 minutes dP is assumed to be. high (> 1000 psi) and water level low (just above top of tube bundle). For this period, neither retention nor scrubbing is assumed and the overall factor is 1.0. For times greater than 4 minutes, the aP decreases to approximately 300 psi and remains constant. for times greater than 4 but less than 32 minutes, retention by the separators is constant and at a maximum. At 32 minutes the separators begin to flood and at 47 minutes the generator is filled. Retention by the separators decreases from the maximum at 32 minutes to zero at 47 minutes. Scrubbing increases with rising water level.

d- The 1 gpm primary to secondary leak is assumed to be split evenly between the steam generators.

29

e. All noble gas activity in. the reactor coolant which is transported to the secondary system via the tube rupture and the primary-to-secondary leakage is assumed to be immediately released to the environment.

f. Case I assumes 30 minute operator action to teminate break flow. The liquid level in the faulted SG remains below the moisture separator. Case 2 assumes 60 minute operator action. The moisture separator begins to flood at 32 minutes and the generator is filled at 47 minutes.

g. The elemental iodine partition factor between the liquid and steam of the intact SG is assumed to be 100. The time dependent partition factor for the faulted SG is presented in Figure III.1-3.

h. Offsite power is lost following reactor trip.

i.. Eight hours after. the accident, the RHR system is assumed to be in opera'tion 'to cool down the plant. Thus, no additional steam release is assumed.

j.

~

~ Neither radioactive decay, during release

~ ~

and transport, nor ground deposition of activity was considered.

~ ~ ~

~

k. Short-term atmospheric dispersion factors (x/g's) for accident analysis and breathing rates are provided in Table III.1-4.

1. Decay constants, average beta and gamma energies and thyroid dose conversion factors are presented in Table III.1-5.

30

OFFSITE THYROID DOSE CALCULATION MODEL Offsite thyroid doses are calculated using the equation Th where (IAR) integrated activity of isotope i released*

during the time interval j in Ci and breathing r ate during time interval j in meter /second offsite atmospheric dispersion factor during time interval j in second/meter (DCF). thyroid dose conversion factor via inhalation for isotope i in rem/Ci thyroid dose via inhalation in rems OFFSITE TOTAL-BODY DOSE CALCULATIONAL MODEL Assuming a semi-infinite cloud of beta and gamma emitters, offsite total-body doses are calculated using the equation:

DTB 0 25Z 5; g (IAR);. (XID).

i j 31

where Integrated activity of isotope i released*

during the j time interval in Ci and offsite atmospheric dispersion factor during time interval j in second/meter E- conservatively assumed to be the sum of the beta and gamma energy for the i isotope in mev/di s.

'TB total-body dose in rems

• No credit is taken for cloud depletion by ground deposition. and radioactive decay during transport to the exclusion area boundary or to the outer boundary of the low-.population zone.

Resul ts Thyroid and Total-Body doses at the Site Boundary and Low Population Zone are presented in Table III.1-6. All doses are within the guidelines of 10CFR100.

32

I TABLE III.1-1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAN GENERATOR TUBE RUPTURE (SGTR)

Source Data

a. Core power level, MWt 1520

b. Steam generator tube 1 leakage, gpm

c. Reac tor -cool ant iodine activi ty:

1..Accident Initiated Spike Initial activity equal to the dose equivalent of 1.0 pCi/gm of I-131 with an assumed iodine spike that increases the rate of iodine release into the reactor coolant by a factor of 500. See Tables III.1-2 and 3.

2. Pre-Accident Spike An assumed pre-accident iodine spike, which has resulted in the dose equivalent of 60 pCi/gm of I-131 in the reactor coolant.

d. Reactor coolant noble gas Based on 1-percent failed I

activity, both cases fuel as provided in Table II I.1-3.

33

TABLE III.1-1 [Sheet 2)

e. Secondary system ini tial Dose equivalent of activi ty O.l pCi/gm of I-131

f. Reactor coolant mass, grams 1.27 x 10

g. Steam generator mass 3.39 x 10 (each), grams

h. Offsite power Lost

i. Primary-to-secondary

!

Case 1 - 30 min 1 eakage duration Case 2 - 60 min

j. Species of iodine 99.9 percent elemental 0.1 percent organic I I. Atmospheric Dispersion Factors See Table III.1-4 III. Activig Release Data

a. Faul ted steam generator

1. Reac tor cool ant discharged See Table III.1.2-2 or 3 to steam generator, lbs.

2. Fl ashed reac tor coolant, See Figure III.1-1 frac tion

3. Iodine attenuation factor See Figure III.1-2 I for flashed fraction of reac tor cool ant 34

TABLE III.1-1 (Sheet 3)

4. Total steam release, See Table III.1.2-2 or 3 lbs

5. Iodine parti ti on fac tor See Figure III.1-3 for the nonf lashed frac tion of reac tor coolant that mixes with the initial iodine activity in the steam genera tor t

6. Location of tube rupture Top of Bundle

b. Intac t steam generator

1. Primary- to-secondary 180 1 ca/age, 1bs/hr

2. Fl ashed reac tor. coolant, frac tion

3. Total steam release, See Table III.1.2-2 or 3 lbs

4. Iodine partition factor 100

5. Isolation time, hrs 35

TABLE I I I.1-2 IODINE APPEARANCE RATES IN THE REACTOR COOLANT {CURIES/SECOND)

FOR A DESIGN BASIS SGTR I-131 I-132 I-133 I-134 I-135 Equi librium Appearance Rates due to Technical Specification Fuel defects 1.88 x 10 4.44 x 10 3.48 x 10 6.14 x 10 4.68 x 10 Appearance Rates due to an Iodine Spike-500X equi librium rates 0.94 2.22 1.74 3.07 2.34

TABLE I II.1-3

.REACTOR COOLANT IODINE AND NOBLE GAS ACTIVITY Nucl ide *Iodine Activity based on 1 pCi/gram of Dose Equiv. I-131 I-131 0.785 pCi/gram I-132 0. 344 I-133 1. 01 I-134 0. 204 I-135 0.787 Noble Gas Activity Based on 1 percent Fuel Defects Xe-131m 1.8 pCi/gram Xe-133m 15 Xe-133 240 Xe-135m 0.41 Xe-135 7.98 Xe-138 0.454 Kr-85m 2.04 Kr-85 6.9 Kr-87 1.18 Kr-88 3.58

• Secondary coolant iodine activity is based on 0.1 pCi/gram of Dose Equivalent I-131 and is therefore 10 percent of these values.

37

TABLE II I.1-4'HORT-TERN ATt10SPHERIC DISPERSION FACTORS AND BREATHING RATES FOR ACCIDENT ANALYSIS Time Site Boundary ~ j Low Population ~ j Breathing ~ j (hours) x/g(Sec/m ) Zone x/g(Sec/m 3 ) Rate (m /Sec) 0-2 48x104 3.47 x 10 4 0-8 3x10~ 3.47 x 10 38

TASLE I II.1-5 ISOTOPIC DATA Decay Constant E E~

DCF~8j Y

~Isoto e (UHr) (Mev/dis) (Mev/di s) (R/ci)

I-131 0.00359 1.49(6)

I-132 0.301 1.43(4)

I-133 0.033 2.69(5)

I-134 0.800 3.73(3)

I-135 0.103 5.60(4)

Xe-131m 0.00245 0.0029 0.165 Xe-133m 0.0128 0.020 0.212 Xe-133 0.00548 0.03 0.153 Xe-135m 2.67 0.43 0.099 XG-135 0.0753 0.25 0.32 Xe-138 2.45 1.2 0.66 Kr-85m 0.158 0.16 0.25 Kr-85 0.00000735 0.0023 0.251 Kr-87 0.547 0.793 , 1.33 Kr-88 0.248 2.21 0.25 39

TABLE 111.1-6 RESULTS OF DESIGN BASIS ANALYSIS Doses (Rem)

Case 1 Case 2

1. Accident Initiated Iodine Spike Site boundary 0-2 hr.)

Thyroid 2.9 91.5 To ta 1 -body 0.31 0.5 Lo w Population Zone (0-8 hr)

Thyroid 0.19 5.7 To ta1 -body 0. 02 0. 03

2. Pre-Accident Iodine S ike Site boundary (0-2 hr)

Thyroid 22.3 273 To ta 1 -body 0.31 0.5 Low Population Zone (0-8 hr)

Thyroid 1.4 17. 1 To ta1 -body 0. 02 0. 03 40

F IGUR E: I II.1-1 O. )000 0.0800 TIME INTERVAL FRACTION I MINUTES) 0 IS 0.055 0.0600 )5-3D 0.020 30-50 '0.0 I O

I-' 5D-60 0.003

)60 0.0 K

4. 0.0400 ID

.O.ozoo 0.0 00 00 0 00 0 0 00 00 00 0 0

0 0 0 0 0 0 0 0 0 0 0 0 P) 0 0 IA 0 0 Ifl TIME (MIN)

BREAK FLOW FLASHING FRACTION

FIGURE:l~ 1 > 2 ZO 30 AO 50 60 TIME t MINUTES)

ATTENUATION FACTOR FOR FLASHEO REACTOR COOLANT 42

l00 50 O

40 a 30 0

20 l0 30 47 TIME (MINUTES)

NORMAL TO BOTTOM S.G.

LEVEL OF MOISTURE FILLED SEP.

FAULTED S.G. PARTITION FACTOR FOR NON FLASHED REACTOR COOLANT 43

III.2 Best Estimate Analytical Assumptions The major assumptions and parameters used in the analysis are itemized in faole III.2-1 and are summarized below.

Source Term Calculations

)he concentrations of nuclides in the primary and secondary system, prior to the accident are determined as follows:

a. The iodine concentrations in the reactor coolant will be based upon preaccident and accident initiated iodine spikes.L ~

i. Preaccident Spike A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine concentration to 8 pCi/gram of Dose Equivalent I-131. (The basis for the spiking factors is presented in Ref. 9.)

ii. Accident Initiated Spike The reactor trip or primary system depressurization associated with the SGTR creates an iodine spike in the primary system which increases the iodine release rate from the tuel to the primary coolant to a value 30L ~ times greater than the release rate corresponding to the maximum equilibrium primary system iodine. concentration of lpCi/gram of Dose Equivalent (O.E.) 1-13l.

The duration of tne spike is assumed to be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Iodine appearance rates in the reactor coolant are presented in Table 2. Doses are calculated for both cases of spiking.

b. The noble gas activity in the reactor coolant is based on 1-percent fuel defects, as provided in Table 3 of Part III.l.

c. Tne secondary coolant activity is based on the O.E. of O.lu Ci/gram of I-131.

d. Iodine at the rupture point is assumed to consist of 100 percent elemental iodine.

The assumption of 1-percent fuel defects for the calculation of noble gas activity is conservative since lgCi/gram D.E. I-131 and I percent defects cannot exist simultaneously. Iodine activity based on I percent defects would be greater than twice the Technical Specification limit.

Dose Calculations The following assumptions and parameters are used to calculate the activity released and the offsite doses following a SGTR.

a. The mass of reactor coolant discharged into the secondary system through the rupture and the mass of steam and/or water released from the intact and faulted steam generators, to the environment is presented in Table III.2-2.

b. The time dependent fraction of rupture flow that flashes to steam and is immediately released to the environment is shown in Figure III.2-1.

c. The time dependent elemental iodine attenuation factor for retention of atomized primary droplets by the moisture separators and dryers and for scrubbing of steam bubbles as they rise from the leak site to the water surface is presented in Figure III.2-2.

Retention by moisture separators and scrubbung are effected by differential pressure (aP) across the ruptured tube and water level. Specifically for the first 5 minutes sP is assumed to be high (550 psi) and water level low (top of tube bundle). For this period, retention and scrubbing are assumed and the overall factor is 1.45. For times greater than 5 minutes the aP decreases to approximately 450 psi and is assumed constant for the duration of the flashing period. for times greater than 5 but less than 22 minutes, retention by the separators is assumed constant and at a maximum. At 22 minutes the separators begin to flood and at 52 minutes the generator and steam line are filled.

Retention by the separators decreases from the maximum at 5 minutes to.zero at 36 minutes. Scruobing increases with rising water level..

d. The I gpm primary to secondary leak is assumed to be split evenly between the steam generators.

e. All noble gas activity in the reactor coolant which is" transported to the secondary system via the tube rupture and the primary-to-secondary leakage is assumed to be immediately released to the environment.

f. The moisture separator begins to flood at 22 minutes and the generator and steam line are filled at 52 minutes.

g. The elemental iodine partition factor between the liquid and steam of the intact SG is assumed to be 5000. The time dependent partition factor for the faulted SG is presented in Figure III.2-3.

h. Offsi te power i s available.

i. 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the accident, the RHR system is assumed to be in opera-tion to cool down the plant. Thus, no additional steam release is assumed.

j.

~

~ Neither radioactive decay, during release and transport, nor ground deposition of activity was considered.

~ ~ ~

~

k. Short-term atmospheric dispersion factors (X/g's) for accident analysis and breathing rates are provided in Table III.2-3.

l. Decay constants, average beta and gamma energies and thyroid dose conver-sion factors are presented in Table 5 of Part III.1.

Offsite Thyroid and Total-8ody Dose Calculational Models See Part III.1 Results Thyroid and total-body doses at the site boundary and low population zone are presented in Table III.2-4. All doses are within the guidelines of 10CFR100.

46

TABLE I I I.2-1 PARAMETERS USED IN THE BEST ESTIMATE EVALUATION THE RADIOLOGICAL CONSEQUENCES OF THE GINNA EVENT I. Source Data

a. Core power 1 evel, MNt 1520

b. Steam generator tube 1 1 eakage, gpm

c. Reactor coolant iodine activi ty:

1. Accident Initiated Spike Initial activity equal to the dose equivalent of 1.0 pCi/gm of I-131 with an assumed iodine spike that increases the rate of iodine release into the reactor coolant by a factor of 30. See Tables III.2-2, III.1-3.

2. Pre-Acc iden t Spike An assumed pre-accident iodine spike, which has resul ted in the dose equivalent of 8 pCi/gm of I-131 in the reactor coolant.

d. Reactor coolant noble gas Based on 1-percent failed fuel activi As provided in Table III.1-3 of Section III.1

e. Secondary system ini tial Dose equivalent of 0.1 pCi/gm activi ty of I-131.

f. Reactor coolant mass, grams 1.27 x 108

g. Steam generator mass (each) grams 3.39 x 10

h. Offsite power Available 47

TABLE I II.2-1 (Continued)

Primary-to-secondary leakage 185 min dura ti on

j. Species of iodine 100 percent elemental II. Atmospheric Dispersion Factors See Table III.2-3 III. Activity Release Data

a. Faul ted steam generator

1. Reactor coolant dis-charged to steam generator, lbs. See Table II.2-2

2. Flashed reactor coolant, See Figure III.2-1 frac tion

3. Iodine attenuation factor See Figure I II.2-2 for flashed fraction of reac tor cool ant

4. Steam and water releases, lbs See Table II.2-2

5. Iodine partition factor for See Figure III.2-3 the nonf lashed fraction of reactor coolant that mixes with the initial iodine activig in the steam generator

6. Location of tube rupture 4 inches above tube sheet

b. Intac t steam generator

1. Primary-to-secondary 180 leakage, lbs/hr

TABLE I II.2-1 (Continued)

2. Fl ashed reac tor cool an t frac tion

3. Total steam release, lbs See Table II.2-2 4~ Iodine partition factor 5000 I sol a ti on time, hrs 21.55

c. Condenser

1. Iodine partition factor 5000 49

TABLE III.2-2 IODINE APPEARANCE RATES IN THE REACTOR COOLANT (CURIES/SECOND)

I-131 I-133 I-134 I-135 Equi librium Appearance Rates due to Technical 1.88 x 10 4.44 x 10 3.48 x 10 6.14 x 10 4.68 x 10 Specification fuel Defects Appearance Rates due to 1.04 x 10 1.84 x 10 1.4 x 10 an Iodine Spike-30X 5.64 x 10 1.33 x 10 equi librium rates

TABLE III.2-3 SHORT-TERM ATMOSPHERIC DISPERSION FACTORS AND 8 REAT HING RATE S FOR ACC I DE WT ANALYSE S Time Site Boundary Low Popul ation Breathing (hours) x/q (Sec/m ) Zone x/g (Sec/m ) Rate (m /sec) 0-2 4.8 x 10 3.47 x 10 0-8 3 x 10 3.47 x 10 8-24 3 x 10 1.75 x 10 Note: x/g's are 10 percent of the R.G. 1.145 values.

51

TABLE I I I. 2-4 RESULTS OF GINNA EVENT ANALYSES Doses (Rem)

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

Thyroid 2.9 To ta 1 -body 0.5 Low Population Zone (0-8 hr)

Thyroid 1.4 To tal -body 0.048

2. Pre Accident S ike Site boundary (0-2 hr)

Thyroid 8.5 To ta 1 -body 0.5 Low Population Zone (0-8 hr)

Thyroid 1.5 To ta1 -body 0..048 52

P F IGuR E: II I 21 O.ZOOO O. l750 O. l500 TIME INTERVAL FRACTION (MINUTES) 0 6 0.!6 O.IZ50 S l7 0.028 0'7 0.0 O

O. IOOO CD K

4 0.0750 x

CA 0.0500

4. I I

I O.OZ50 I I

0.0 O 0 o O 0 0 o lA O lA 0 tA Al O

lA ~

lA r

O CO EV lA TIME ( MIN)

BREAK FLOW FLASHING FRACTION FOR THE GINNA EVENT 53

10 9

8 IO I5 20 30 Tll4E I MlNUTES)

ATTENUATION FACTOR FOR FLASHED REACTOR COOLANT FOR THE GlNNA EVENT 54

5000 I

I I

I I

I a: 1000 I O I f

I I

I O I f-. I F- I I

I 100 I I

I I

I I

I I

I 10 ZO 30 60 TIME I MlNUTES)

FAULTED S.G. PARTIT10N FACTOR FOR'HE GINNA EVENT, I

55

IV.

SUMMARY

AND CONCLUSIONS The potential environmental consequences of a steam generator tube failure at the R. E. Ginna nuclear power plant were evaluated in order to demonstrate

~ ~

~ ~

that the Standard Technical Specifications limit on primary coolant activity

~ ~ ~

is acceptable. The mass releases during a design basis event, i.e. a double ended rupture of a single tube, were conservatively calculated using the com-puter code LOFTRAN. For these analyses, the sequence of recovery actions initiated by the tube failure were assumed to be completed on a restricted time scale. Two cases were considered: a) 30 minute recovery, and b) 60 min'ute recovery. The effect of steam generator overfil1 on radiological was also considered. Mass releases during the design basis event'eleases were used with conservative assumptions of coolant activity, meteorology, and attenuation to estimate an upper bound of site boundary and low population zone exposures.

The mass releases from the January 25, 1982 steam generator tube failure at Ginna were also calculated from results presented in reference 2. These releases were used with the Standard Technical Specification limit on initial coolant activity and a more realistic meteorology to evaluate potential doses on a more realistic basis.

Results of the design basis analyses indicate that the conservative site boundary and low population zone exposures from a steam generator tube failure are within 10CFR100 limitations with the Standard Technical Specification limit on initial coolant activity. Estimates of the potential radiological releases from a more realistic event with the same initial coolant activity demonstrate that the design basis analysis is very conservative. Conse-quently, the Standard Technical Specification limit on coolant activity are sufficient to ensure that the environmental consequences of a steam generator tube failure at the R. E. Ginna plant will be within acceptable limits.

56

REFERENCES

1. L. A. Campbell, "LOFTRAN CODE DESCRIPTION", WCAP-7878 Rev. 3, January (1977).

2. E. C. Volpenhein, "ANALYSIS OF PLANT RESPONSE DURING JANUARY 26, 1982 STEAN GENERATOR TUBE FAILURE AT THE R. E. GINNA NUCLEAR POWER PLANT",

Westinghouse Electric Co., October (1982).

3. WESTINGHOUSE OWNERS GROUP EMERGENCY RESPONSE GUIDELINES SElfINAR, September 1981.

4. NRC Standard Review Plan 15.6-3, Rev. 2, "Radiological Consequences of a Steam Generator Tube Failure", Ju'ly, 1981.

5. NRC NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident", Postma, A.K., Tam, P.S., Jan. 1978.

6- NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential

. Accident Consequence Assessments at Nuclear Power Plants", August, 1979.

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

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

9. Lutz, R. J., "Iodine and Cesion Spiking Source Terms for Accident Analysis," MCAP-9964, Rev. 1, July 1981.

57