ML17256A500

Proposed Tech Spec 3.1.4 Re Max Coolant Activity

ML17256A500
Person / Time
Site:	Ginna
Issue date:	02/18/1983
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17256A498	List: ML17256A497 ML17256A499 ML17256A500
References
NUDOCS 8302230367
Download: ML17256A500 (79)

Text

Attachment A

1.

Replace Technical Specification pages 3.1-21 and 3.1-24 with the enclosed revisions.

8gOP2303b7 8300244 PDR ADOCK o PDR P

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3.1.4 Maximum Coolant Activit S ecifications:

3.1.4.1 Whenever the reactor is critical or the reactor coolant average temperature is greater than 500 F:

0 a ~

The total specific activity of the reactor coolant shall not exceed 84/E -pCi/gm,. where E is the average beta and gamma energies per disintegration in Mev.

b.

The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 1.0 pCi/gm.

c.

The I-131 equivalent of the iodine activity on the secondary side of a steam generator shall not exceed O.l pCi/gm.

3.1.4.2 If the limit of 3.1.4.l.a is exceeded, then be sub-critical with reactor coolant average temperature less than 500 F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

a

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If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.l.b but is less than the allowable limit shown on Figure 3.1.4-1, operation may continue for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> provided that the cumulative operating time under these circumstances does not exceed 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any consecutive 12-month period.

If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.l.b for more than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any consecutive 6-month period, 3.1-21 Amendment No.

Attachment B

The potential environmental consequences of a steam generator tube failure at the R.

E. Ginna nuclear power plant were evaluated in order to demonstrate the Standard Technical Specification limit on primary coolant activity is acceptable.

Results of the design analysis indicate that the conservative site boundary and low population zone exposure from a steam generator tube failure are within 10 CFR 100 limitations with Standard Technical Specifica-tion limits on initial coolant activity.

Therefore, the Standard Technical Specification limit on coolant activity is sufficient to ensure that the environmental consequences of a steam generator tube failure will be within acceptable limits.

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) I.

INTRODUCTION

'I 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'o 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 resulted, if the accident had occurred with coolant activity limits established

. in the standard technical specifications.

g 250 0

200 Q

z I

150 8

+ 100 I

I-z 50 8

0

~

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BLE ACCEPTA OPERATION UNACCEPTABLE OPERATION 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMALPOlNER I-131 FlGURE 3-1.4-1 Equivalent Reactor Coolant Specific Activity Limit Versus

,Percent of Rated Thermal Power 3.1-24 Amendment No.

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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 ere 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:

H 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.

~I C.

The decrease in'CS pressure due to continued loss of reactor coolant inventory leads to a reactor trip signal on low pressurizer pressure or 3'vertemperature 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,

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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 (MSIV) provide this capability.

In the event of a failure of the MISV for the faulted steam generator, the MSIV for the intact steam generator and the turbine stop'alve 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 fo'r 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~rimary pressure is reduced to a value equal to the faulted steam genera-l +

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.

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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.

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'his 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.l.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.

4 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.

~

0 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.

Steam1ine volume is not considered in calculating the time of steam generator over-fill 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 Westinghouse 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-

'ional 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 carryover 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.

With 30 minute =operator action to terminate break flow,

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TABLE II.1.2-1 DESIGN BASIS ACCIDENT SEl}UENCE OF'YENTS Event Manual (0)

Time (Sec)

Automatic (A) 30 Min Recovery 60 Min Recovery

'l Tube Failure Reactor Trip Condenser Lost SI Signal Feedwater Isolation AFM Initiation AFW Throttled to Faulted SG Isolation of Faulted SG Steam Dump RCS Depressurization SG Overfill SI Terminated Break Flow Terminated RHR Cooling A==

0 I

0 0

27 27 127 134 187 187 1800(1 )

1SOO<<)

lsoo(1) 1800(1) 1800(1) 28800 27 27 127 134 187 187 3600(1) 3600(1) 3600(1) 281 0 3600(1 )

3600(1) 28800

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

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I TABLE II.1.2-2 MASS RELEASES DURING A DESIGN BASIS SGTR:

30 MINUTE RECOVERY Flow (ibm) 0-TTRIP Time Period TTRIP-TTBRK TTBRK-2 2-TRHR Ruptured SG:

- Condenser

- Atmosphere

- Feedwater 27820 0.0 32605 0.0 32640 0.0 0.0 0.0 0.0 0.0 21 480 21 480 Intact SG:

- Condenser

- Atmosphere

- Feedwater 27380 0.0 37170 0.0 23050 13370 0.0 144650 206200 0.0 470000 487600 Break Flow 3325 100648

'.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, Flow (ibm)

Time Period 0-TTR IP TTRIP-TMSEP-TSGOF-TTBRK-2 2-TRHR TMSEP TSGOF TTBRK

- Atmosphere

- Feedwater 0.0 32605 Ruptured SG:

- Condenser 27820 0.0 0.0 0.0 0.0 33570 4830 0.0 0.0 43171 0.0 0.0 0.0 0.0 0.0 0.0 Intact SG:

- Condenser

- Atmosphere

- Feedwater 27380 0.0 37170 0.0 23370 13700

'.0 1390 1390 0.0 390 380 0.0 67970 129600 0.0 501100 518700 Break Flow 3325 107742 48070 43171 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

.0 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 faulte'd 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.l;2-5 PRESSURIZER WATER VOLUME FIGURE II.1.2-6 FAULTED SG STEAM FLOW FIGURE II. 1. 2-7 BREAK FLOW FIGURE II.1.2-8 BREAK FLOW FLASHING FRACTION II. 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

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FAULTED STEA51 GEHERATOR HATER VOLU)1E.

11

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1250.0 1000.0 750. 00 500.00 300.00 CD C)

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AJ TlHC (MIN)

CD CD CD CD CD ICl CD CD CD

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REACTOR COOLANT SYSTdl PRESSURE.

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FAULTED STEAH GENERATOR PRESSURE.

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~ 300.00 l

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CD CD CD AJ CD CD CD m

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C7 CD CD CD CD CD lA CD CD CD IO FIGURE II. 1 2-4.

REACTOR COOLANT AVERAGE TEI1PERATURE

800.00 600.00 500.00 m

I au 400.00 Cl) 300.00 l

200.00 300.00 0;0 Cl Cl Cl Cl Cl Cl AJ-Cl m

TIME (HIM)

Cl FIGURE II.1.2-S.

PRESSURIZER HATER VOLUt'lE.

0.1750 0.1500 O

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0.0750 CD la 0.0500 0.0250 0.0 CD ED CD Cl CD TINE (MlN)

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FAULTED STEAM GENERATOR STEAN FLOW.

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100.00 15.000 CD

L 50.000 25.000 0.0

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CD CD CD CD If1 FIGURE II.1.2-7.

PRIl1ARY-TO-SECONDARY LEAKAGE.

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AJ C7 Cl m

TIME (M1N)

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CD ED ED CD CD CD ED CD FIGURE II.1-2-8.

BREAK FLOW FLASHING FRACTION.

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TABLE II.2-1 GINNA SEQUENCE OF EVENTS Event Manual (0)

Automatic (A)

Actual Time (sec)

Simulated Tube Failure Reactor Trip Condenser Lost SI Signal Feedwater Isolation AFW Initiated AFW Throttled to Faulted SG Isolation of Faulted SG Steam Dump RCS Depressurization SG Overfill SI Terminated Break Flow Terminated RHR Cooling 0

0 0

0 0.

0 0

0 182 4500 190 192 220 410 890 770 2700 4310 10800 77580 0

182 4500 198 198 239 410 530 530 2700 3130 4310 10800 77580 includes steamline volume 19

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event described in section II.1.1.

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 4

feedwater and safety 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-tri p 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 within 15 minutes of tube failure by closing the associated MSIV.

Continued auxiliary feedwater flow to the intact steam gene-

'ator 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 PORY 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 Flow (1bm)

Time Period 0-TTRIP TTRIP-TNSEP-TSGOF*-2 2-TTBRK TTBRK-TNSE P TSGOF*

TRHR Faulted SG:

- Condenser

- Atmosphere

- Feedwater 162100 16900 0

0 163400 46800 0

0 130442 0

0 105684

. 0

~

~ 0 Intact SG:

- Condenser

- Atmosphere

- Feedwater I

160100 288000 25200 14500 I

0 0

0 23870 171700 52300 0

,89700 0

54743 53008 0

978387 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

'f000. 0 6000.0 S.G.

AND STEAt'1LIilE VOLUflE 5000.0 S.G.

VOLUtlE m 4000.0 I

~~ 3000.0 2000.0 1000.0 0.'0 O

O OOO ldll OO O

C7 O

Ill OO TlHE'MIN)

O C)

V1 AJ C)O O

aoO

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FIGURE II'-1., CALCULATED FAULTED STEAfl GENERATOR MATER VOLUHE DURING THE GINNA EVENT.

22

2300.0 2250.0 2000.0

~

1750.0 1500. 0 1250.0 G

G C

G G

1000.0

.G G

G 750.00 500.00 300.00 D

O O

l/I AJ DD OO AJ OO OI/i O D D D

~

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Ifl O TlHE <MlN)

FIGURE II.2-2.

REACTOR COOLAHT SYSTEsl PRESSUPE DURING THE GIHiNA EVEiAT.

23

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FIGURE II.2-3.

FAULTED STEAh GENERATOR PRESSURE DURING THE GINNA EVENT.

24

o.zoao

0. 5500 0

$250

~ O.tNN 0 0750 0 0500 0 0259 0 0 8

8 T1m <atm FIGURE II.2-4.

CALCULATED BREAK FLOW FLASHI'AG FRACTION DUR!NG TIIE GIHl<A EVEi<T.

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.

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III.

ENVIRONMENTAL CONSE(}UENCES ANALYSIS Introduc tion 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 primary coolant enter the steam generator, via'he ruptured

tube, and are released to the atmosphere through the steam generator safety or power ope'rated relief val ves.

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 break f'1ow between the steam generator 1iquid 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.l.

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 fo11owing sections.

III.1 DESIGN BASES ANALYTICALASSUMPTIONS, The major assumptions and parameters used in the analysis are itemized in Table I I.l-l and are summarized below.

E 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 1pCi/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 I

rates in the reactor coolant are presented in Table III.1-2.

Doses are:

I 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.

'I The assumption of 1 percent fuel defects for the calculation of noble gas activity is conservative, since luCi/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 D.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

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 fac tor 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 oy moisture separators and scrubbing are effected by differential pressure (dP) across the ruptured tube and water level.

Specifically for the first 4 minutes aP 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 dP 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 seco'ndary 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 1 assumes 30 minute operator action to terminate break flow.

The liquid level in the faul ted 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 111.1-4.

l.

Decay constants, average beta and gamma energies and thyroid dose conversion factors are presented in Table 111.1-5.

30

0 OFFSITE THYROID DOSE CALCULATION MODEL Offsite thyroid doses are calculated using the equation:

where DIA

--g DCF.

g (IAR)..

(BR).

(X/0).

(IAR) integrated activity of isotope i released*

during the time interval j in Ci and (BR).

breathing rate 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 CALCULATIONALMODEL Assuming a semi-infinite cloud of beta and gamma emitters, offsite total-body doses are calculated using the equation:

DTB

= 0.25+

5 g

(IAR)

(2/0) 1 j

31

I I

where (IAR)..ij Integrated activity of isotope i released*

during the j time interval in Ci and (x/g).

offsite atmospheric dispersion factor during time interval j in second/meter E.

conservatively assumed to be'he sum of the beta.and gamma energy for the i isotope in mev/dis.

TB total-body dose in rems l a 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 lo'w-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

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

Source Data a.

Core power level, MWt b.

Steam generator tube

leakage, gpm c.

Reactor coolant iodine ac tivity:

hI 1.

Accident Initiated Spike 1520 1

Initial activig 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 cool ant.

d.

Reactor coolant noble gas activity, both cases Based on 1-percent failed fuel as provided in Table III.1-3.

33

TABLE III.1-1 (Sheet 2) e'.

Secondary system initial ac tivity Dose equivalent of O.l pCi/gm of I-131 f.

Reactor coolant mass, grams g.

Steam generator mass (each),

grams 1.27 x 10 3.39 x 10 h.

Offsite power Lost Primary-to-secondary leakage duration Species of iodine 99.9 percent elemental 0.1 percent organic Case 1 - 30 min Case 2 - 60 min II.

Atmospheric Dispersion Factors See Table III.1-4 III.

Activity Release Data a.

Faul ted steam generator 1.

Reactor coolant discharged to steam generator, lbs.

See Table III.1.2-2 or 3 2.

Fl ashed reac tor coolant, frac tion See Figure III.1-1 3.

Iodine attenuation factor for flashed fraction of reac tor coolant See Figure III.1-2 34

TABLE III.1-1 (Sheet 3) 4.

Total steam

release, lbs See Table III.1.2-2 or 3 5.

Iodine parti tion factor for the nonflashed frac tion of reac tor coolant that mixes with the initial iodine activity in the steam genera tor See Figure III.1-3

,1 6.

Location of tube rupture Top of Bundle b.

Intact steam genera'tor 1.

Primary-to-secondary l ca/age, lbs/hr 180 2.

Fl ashed reac tor cool ant, frac tion 0

3.

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

Iodine partition factor 100 5.

Isolation time, hrs 8

35

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

FOR A DESIGN BASIS SGTR I-131 1-132 1-133 I-134 1-135

'quilibrium Appearance Rates due to Technical Specification-Fuel defects 1.88 x 10 4.44 x 10 3.48 x 10 6.14,'

10 4.68 x 10-3 Appearance Rates due to an Iodine Spike-500X equi librium rates 0.94

2. 22 1.74 3.07 2.34

TABLE III.1-3 REACTOR COOLANT IODINE AND NOBLE GAS ACTIVITY Hucl ide

• Iodine Activity based on 1 pCi/gram of Dose Equiv. I-131 I-131 I-132 I-133 I-134 I-135 0.785 pCi/gram

0. 344 1.01 0.204 0.787 Noble Gas Activity Based on 1 percent Fuel Defects Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-138 Kr-85m Kr-85 Kr-87 Kr-88 1.8 pCi/gram 15 240 0.41 7.98 0.454 2.04 6.9 1.18 3.58

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

37

TABLE III.1-4 SHORT-TERN ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES FOR ACCIDENT ANALYSIS Time Site Boundary

~ j (hours) x/g(Sec/m

)

Low Population

~ j Zone x/g(Sec/m

)

3 Breathing

~ j Rate (m /Sec) 0-2 4.8 x 10 4 3.47 x 10 4 0-8 3xl0~

3.47 x 10

TABLE III.1-5 ISOTOPIC DATA Decay Constant Iso tope (1/Hr)

E Y

(Mev/dis)

E~

(Hev/di s)

DCFL81 (R/ci)

I-131 I-132 I-133 I-134 I-135 0.00359 0.301 0.033 0.800 0.103 1.49(6) 1.43(4) 2.69(5) 3.73(3) 5.60(4)

Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-138 0.00245 0.0128 0.00548 2.67 0.0753 2.45 0.0029 0.020 0.03 0.43 0.25 1.2 0.165 0.212 0.153

0.099 0.32 0.66 Kr-85m Kr-85 Kr-87 Kr-88 0.158 0.00000735 0.547 0.248 0.16 0.0023 0.793 2.21 0.25 0.251 1.33

, 0.25 39

a I

TABLE III.1-6 RESULTS OF DESIGN BASIS ANALYSIS Doses (Rem)

Case 1

Case 2

r 1.

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

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

Thyroid Total-body 0.19

0. 02 5.7

0. 03 2.

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

Thyroid Total-body 22.3 0.31 273 0.5 Low Population Zone (0-8 hr)

Thyroid Total-body 1.4

0. 02

17. 1 0.03 40

0.0800 0.0600 TIME INTERVAL I M IHUTES) 0-l5 I5-30 30-50 50-60 060 FRACTION 0.055 0.020 O.OI 0.003 0.0 0.0400 I

I 0.0 O

O OOO O

OOO ON OO O

Pl OOO0 oOO O

tA OoO O

O O

o o

Q O

O tA TIME (MIN)

BREAK FLOW Fl ASHING FRACTION.

IO 5

O I

CJ O

3 0

to ZO 30 40 TIME (MINUTESI 50 ATTENUATION FACTOR FOR FLASHEO REACTOR. COOLANT 42

1 I

t00 50 CC C) 40 a

30 I

I 20 IO 0

30 T1ME l MlNUTES)

NORMAL LEVEL TO BOTTOM OF MOlSTURE SEP.

S.G.

F1LLED FAULTED S.G. P AR7 17I0 N FACTOR FOR NON FLASHED REACTOR COOLANT 43

III.2 Best Estimate Anal tical Assumptions The major assumptions and parameters used in the analysis are itemized in l aole III.2-1 and are summarized below.

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

a.

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

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 I

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.) 1-131.

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 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 Tan)e 3 of Part III.l.

c.

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

d.

Iooine at the rupture point is assumed

'to consist of 100 percent elemental iodine.

44

I

The assumption of I-percent fuel defects for the calculation of noble gas activity is conservative since lpCi/gram D.E.

I-131 and I percent defects cannot exist simultaneously.

Iodine activity based on 1.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.

I b.

The time dependent fraction of rupture flow that flashes to steam and is immediately released to the environment is shown in Figure II.I.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.

I

)

I I

I

~

~

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.

Offsite power is 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 aimospheric dispersion factors (X/g's) for accident analysis and breathing rates are provided in Table III.2-3.

1.

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-Body Dose Calculational Models See Part III.1 Resul ts 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 III.2-1 PARAMETERS USED IN THE BEST ESTIMATE EVALUATION THE RADIOLOGICAL CONSE(}UENCES OF THE GINNA EVENT I.

Source Data a.

Core power level, tQt

~

b.

Steam generator tube

leakage, gpm c.

Reactor coolant iodine activi g 1520 1

I 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-Accident Spike An assumed pre-accide'nt iodine

spike, which has resulted in the dose equivalent of 8 pCi/gm of I-131 in the reactor coolant.

d.

Reactor coolant noble gas activi Q Based on 1-percent failed fuel As provided in Table III.1-3 of Sec tion III.l e.

Secondary system initial ac tivity f.

Reactor coolant mass, grams g.

Steam generator mass (each) grams h.

Offsite power Dose equivalent of. 0.1 pCi/gm of I-131.

1.27 x 10 3.39 x 10 Available 47

J I

~

V TABLE III.2-1 (Continued) i.

Primary-to-secondary 1 eakage duration Species of iodine 185 min 100 percent elemental II. Atmospheric Dispersion Factors See Table III.2-3 III.

Ac tivity Release Data a.

Faul ted steam generator 1.

Reactor coolant dis-charged to steam generator, lbs.

See Table II.2-2 2.

Fl ashed reac tor cool ant, frac tion 3.

Iodine attenuation factor for flashed fraction of

.: reactor coolant 4.

Steam and water releases, lbs I

5.

Iodine partition factor for the nonflashed frac tion of reactor coolant that mixes with the initial iodine activity in the steam generator 6.

Location of tube rupture See Figure III.2-1 See Figure III.2-2 See Table II.2-2 See Figure III.2-3 4 inches above tube sheet b.,

Intact steam generator 1.

Primary-to-secondary

leakage, lbs/hr 180

I l

I

'TABLE III.2-1 (Continued)

. ~

2.

Flashed reac tor coolant frac tion 3.

Total steam release, lbs 4.

Iodine partition factor 5.

Isolation time, hrs 0

See Table II.2-2 5000 21.55 c.

Condenser 1.

Iodine partition factor -,-

5000

I I

I I

I

TABLE III.2-2 IODINE APPEARANCE RATES IN THE

- REACTOR COOLANT (CURI'ES/SECOND)

I-131 I-132 I-133 I-134 I-135 Equi 1 ibrium Appearance Rates due to Technical Specification Fuel Defects 1.88 x 10 4.44 x 10 3.48,x 10 4.68 x 10 Appearance Rates due to an Iodine Spike-30X equi librium pates 5.64 x 10 1.33 x 10 1.04 x 10 1.84 x 10 1.4 x 10

TABLE III.2-3 SHORT-TERM ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES FOR ACCIDENT ANALYSES Time (hours)

Si te Boundary [6j x/Q (Sec/m

)

Low Population L6j Zone x/Q (Sec/m

)

Brea thing Rate (m /sec) 0-2 0-8 4.8 x 10 3 x.10 3.47 x 10 3.47 x 10 8-24 3 x 10 1.75 x 10

\\

Note:

x/Q's are 10 percent of the R.G.

1.145 values.

51

TABLE III.2-4 I

RESULTS OF GINNA EVENT ANALYSES l.

Accident Initiated Iodine Spike Doses (Rem)

Site boundary (0-2 hr)

Thyroid To ta1 -body 2.9 0.5 Low Population Zone (0-8 hr)

Thyroid Total-body 1.4 0.048 2.

Pre Accident Spike Site boundary (0-2 hr)

'hyroid Total-body 8.5 0.5 Low Population Zone (0-8 hr)

Thyr oid To tal -body 1.5 0.048.

52

E AUGUR E:

I I I.2-1 0.2000 O. I750 O. I500 O.I250 TIME INTERVAL

( IAlNUTE5) 0-6 5"I7

>l7 FRACTION O.IG 0.029 0.0 O. IOOO O

0 0.0750 0.0500 4.

0.0250 1

I I

I I

0.0 Oe O

M O

O O

IA O

O O

lA O

O lA CO TIME (MINI BR EAK Fl QW FLASHING FR ACTI0 N FOR THE GINNA EYENT 53

10 9

8 S

C) ~

I0 4

OI-l4 I

0 IO IS 20 TIME (MINUTES) 25 30 ATTENUATION FACTOR FOR FLASHED REACTOR COOLANT FOR THE GINNA EVENT 54

t 4

A

5000 a:

IOOO O

a f-.

F-

~

too I

I I

,i I

I I

t I

I I

I I

I I

I t

I II, I

II I

I t

I 0

IO ZO 30 50 60 TlME I MlNUTE5)

FAULTED S.G. PARTlTION FACTOR FOR'HE GINNA EVENT 55

l~

t

)

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

'nded rupture of a single tube, were conservatively calculated using the com-puter co'de 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 minute recovery.

The effect of steam generator overfill on radiological'eleases was also considered.

Mass releases during the design basis event 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

're within 10CFR100 limitations w'ith 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

v li I

REFERENCES 1.

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

2.

E.

C. Volpenhein, "ANALYSIS OF PLANT RESPONSE DURING JANUARY 26, 1982 STEAM GENERATOR TUBE FAILURE AT THE R.

E.

GIHNA NUCLEAR POWER PLANT",

Westinghouse Electric Co., October (1982).

3.

WESTINGHOUSE OWNERS GROUP EMERGENCY RESPONSE GUIDELINES SEMINAR, September 1.981.

4.

NRC Standard Review Plan 15.6-3, Rev.

2, "Radiological Consequences of a Steam Generator Tube Failure", July, 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.

r 6.

NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants",

August, 1979.

7.--HRC Regulatory-Guide 1.4, Rev.

2, "Assumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for Pressurized Water 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," WCAP-9964, Rev.

1, July 1981.

57

I~N b