ML20133H845
| ML20133H845 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/31/1985 |
| From: | Brennan J, Rubin K WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17254A579 | List: |
| References | |
| WCAP-10885, NUDOCS 8510180142 | |
| Download: ML20133H845 (62) | |
Text
___ _
WESTINGHOUSE CLASS 3 WCAP-10885 ANALYSIS OF POTENTIAL RADIOLOGICAL CONSEQUENCES FOLLOWING A STEAM GENERATOR TUBE RUPTURE AT THE R. E. GINNA NUCLEAR POWER PLANT USING LOFTTR1 J. Brennan K. Rubin JULY 1985 Nuclear Safety Department Westinghouse Electric Corporation Nuclear Energy Systems l
P.O. Box 355 Pittsburgh, Pennsylvania 15230
(#
1985 by Westinghouse Electric Corporation 90238:10/082185 8510180142 851009 PDR ADOCK 05000244 P
WESTINGHOUSE CLASS 3 TABLE OF CONTENTS Pages 1.
INTRODUC110N 1
4 II.
THERMAL HYDRAULIC ANALYSIS 2
A.
Design Basis Accident 2
B.
Conservative Assumptions 3
C.
Operator Action Times 6
i D.
Transient Description - Case 1 12 1
E.
Transient Description - Case 2 23 F.
Mass Releases 33 i
III.
RADIOLOGICAL CONSEQUENCES ANALYSIS 42 IV.
CONCLUSION 57 V.
REFERENCES 58 s
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WESTINGHOUSE CLASS 3 1.
INTRODUCTION An evaluation of the potential radiological consequences due to a steam generator tube rupture (SGTR) event has been performed for the R. E. Ginna nuclear power plant to demonstrate that the offsite radiation doses will be less than the allowable guidelines based on the Standard Technical Specification limit on primary coolant activity.
i i
A design basis steam generator tube rupture was analyzed for Ginna using the methodology. developed in WCAP-10698 (reference 1) and the supplement to i
WCAP-10698 (reference 2). Two single failure cases were considered to determine which is the most limiting single failure for Ginna with respect to radiological consequences. The two cases examined were:
_?' c Case 1 i
i Case 2 l Plant response to the event was modeled using the LOFTTR1 computer code with conservative assumptions of break size and location, condenser availability and initial secondary water mass in the faulted steam generator. The analysis methodology includes the simulation of the operator actions for recovery from a steam generator tube rupture based on the Westinghouse Owners Group Emergency Response Guidelines, which are the basis for the Ginna Emergency Operating Procedures. The mass releases were calculated with the LOFTTR1 program from the initiation of the event until termination of the break flow.
For the time period following break flow termination, steam releases and j
feedwater flows from the intact and faulted steam generators were determined from a mass and energy balance using the calculated RCS and steam generator conditions a. the time of leakage termination. The mass-releases for both cases were used to determine the radiation doses at the exclusion area boundary and low population zone assuming that the primary coolant activity is at the Standard Technical Specification limit prior to the accident.
90238:1D/081985 1
WESTINGHOUSE CLASS 3 II. THERMAL HYDRAULIC ANALYSIS Integrated mass releases to the atmosphere and condenser during a steam generator tube rupture event were calculated for various time periods during the accident. This section includes the methods and assumptions used to model the SGTR event and calculate the mass releases, as well as the sequence of events for the recovery.
1 A.
Design Basis Accident The accident modeled is the complete severance of a steam generator tube located at the tube sheet on the cold leg side.
It was also assumed that loss of of fsite power occurred at the time of reactor trip, and the worst rod was assumed to be stuck at reactor trip.
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4 1
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WESTINGHOUSE CLASS 3 B.
Conservative Assumptions Plant' responses and mass releases from the intact and faulted steam generator prior to break flow tennination were calculated using LOFTTRI.
While modeling the SGTR event the following assumptions were made:
1.
Reactor Trip on Overtemperature delta-T i
s 2.
Power g
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Pressurizer Water Level
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WESTINGHOUSE CLASS 3 4.
Steam Generator Secondary Mass a,c.
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S.
Break Location og 3
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6.
Reactor Trio Delav 4
7.
Turbine Trio Delay I
i.
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WESTINGHOUSE CLASS 3 8.
Steam Generator Relief Valve Pressure Setpoint 44 9.
Pressurizer Pressure for SI Initiation c,c
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4 10 Leakage after Overfill a
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- 11. Flashina Fraction qn 4
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WESTINGHOUSE CLASS 3 C.
Operator Action Times i
In the event of an SGTR, the operator is required to take actions to stabilize the plant and terminate the primary to secondary leakage. An evaluation has been performed (reference 1) to establish the operator a
action times for use in the analysis of a design basis SG1R event.
The operator actions which are required for recovery f rom an SGTR and the available data on the times to perform these actions have been reviewed.
The available data on operator action times for an SGTR includes information which has been obtained from reactor plant simulator studies as well as plant data from five actual SGTR events.
Using this data, operator action times have been established which are considered to be appropriate for a design basis SGTR event.
These operator action times will be used as input for the analysis of the design basis SGTR event.
The major operator action for SGTR recovery which are included in the E-3 guideline of the Westinghouse Owners Group Emergency Response Guidelines were explicitly modeled in the analysis. The operator actions modeled include identification and isolation of the ruptured steam generator, cooldown and depressurization of the RCS to restore inventory, and termination of SI to stop primary to secondary leakage. These operator actions are described below.
1.
Identify the ruptured steam generator (step 2).
4 High secondary side activity, as indicated by the steam generator blowdown line radiation monitor or air ejector radiation monitor, typically will provide the first indication of an SGTR event.
The ruptured steam generator can be identified by high activity in the corresponding steam generator blowdown line, main steamline, or water sample.
For an SGTR that results in a high power reactor trip, the steam generator water level i
90238:10/081985 6
WESTINGHOUSE CLASS 3 will decrease of f-scale on the narrow range for both steam generators.
The auxiliary feedwater ( AFW) flow will begin to refill the steam generators, typically distributing approximately equal flow to both steam generators. Since primary to secondary leakage adds additional inventory which accumulates in the ruptured steam generator, level will return to the narrow range in that steam generator significantly earlier and will continue to increase more rapidly. This response provides confirmation of an SGTR event and also identifies the ruptured steam generator.
2.
Isolate the ruptured steam generator from the intact steam generator and isolate feedsater to the ruptured steam generator.(steps 3 and 4).
Once a tube rupture has been identified, recovery actions begin by isolating steam flow from and stopping feedwater flow to the ruptured steam generator.
In addition to minimizing radiological releases, this also reduces the possibility of filling the ruptured steam generator with water by 1) minimizing the accumulation of feedwater flow and 2) enabling the operator to establish a pressure differential between the ruptured and intact steam generators as a necessary step toward terminating primary to secondary leakage.
In the guideline for steam generator tube rupture recovery in the ERGS, the operator is directed to maintain the level in the ruptured steam generator between just on span and 50% on the narrow range instrument.
Reference 1 assumed that the ruptured steam generator would be isolated when level in the steam generator reached midway between these points.
For Ginna it was conservative to use 33 percent of level for isolation.
If the time to reach 33 percent narrow range level was less than 10 minutes, then 10 mir.utes was used as the isolation time. The ruptured steam generator was assumed to be isolated at 33 percent narrow range level or at 10 minutes, whichever was longer.
3.
Cool down the Reactor Coolant System (RCS) using the intact steam generator (step 14).
After isolation of the ruptured steam generator, the RCS is cooled as rapidly as possible to less than saturation at the ruptured steam 9023B:10/081985 7
l
WESTINGHOUSE CLASS 3 generator pressure by dumping steam f rom only the intact steam generator.
This ensures adequate subcooling in the RCS af ter depressurization to the ruptured steam genarator pressure in subsequent actions. With offsite power available, the normal steam dump system to the condenser will provide suf ficient capacity to perform this cooldown rapidly.
If offsite power is lost, the RCS is cooled using the power-operated relief valve (PORV) on the intact steam generator since neither the steam dump valves nor the condenser would be available.
.t is noted that RCS pressure will decrease during the cooldown as shrinkage of the reactor coolant expands the steam bubble in the pressurizer.
4.
Depressurize the RCS to restore reactor coolant inventory (steps 17 or 18).
When the cooldown is completed, SI flow will increase RCS pressure until break flow matches SI flow.
Consequently, SI flow must be terminated to stop primary to secondary leakage.
However, adequate reactor coolant inventory must first be assured. This includes both sufficient reactor coolant subcooling and pressurizer inventory to maintain a reliable pressurizer level indication after SI flow is stopped. Since leakage from the primary side will continue af ter SI flow is stopped until RCS and ruptured steam generator pressures equalize, an " excess" amount of inventory is needed to ensure pressurizer level remains on span. The
" excess" amount required depends on RCS pressure and reduces to zero when RCS pressure equals the pressure in the ruptured steam generator.
To reduce break flow and establish sufficient pressurizer level, RCS pressure is decreased by opening the pressurizer PORV.
5.
Terminate SI to stop primary to secondary leakage (steps 21-23).
The previous actions will have established adequate RCS subcooling, secondary side heat sink, and reactor coolant inventory following an SGTR to ensure that SI flow is no longer needed. When these actions have been completed, SI flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary leakage.
Primary to secondary leakage will continue after SI flow is stopped until RCS pressure and 9023B:10/081985 8
I I
WESTINGHOUSE CLASS 3 ruptured steam generator pressures equalize.
Charging flow, letdown, and pressurizer heaters will then be controlled to prevent repressurization of the RCS and reinitiation of leakage into the ruptured steam generator.
Since these major recovery actions will be modelled in the SGTR analysis, it is necessary to establish the times required to perform these actions.
Although the intermediate steps between the major actions will not be explicitly modelled, it is also necessary to account for the time required to perform the steps.
It is noted that the total time required to complete the recovery operations consists of both operator action time and system, or plant, response time.
For instance, the time for each of the major recovery operations (i.e., RCS cooldown, RCS depressurization, etc.) is primarily due to the time required for the system response, whereas the operatar action time is reflected by the time required for the operator to perform the intermediate action steps.
O The operator action times to initiate RCS cooldown, RCS depressurization and safety injection termination were developed in reference 1 and are listed in Table 11.1.
In addition to the operator action times developed in reference 1, Rochester Gas and Electric supplied the operator action times associated i
with recovering from the single failures (Reference 3).
The times associated with performing these operator actions are listed in Table 11.2.
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WESTINGHOUSE CLASS 3 TABLE 11.1 OPERATOR ACTION TIMES FOR DESIGN BASIS SGTR ANALYSIS Y
Action Identify and isolate ruptured SG Operator action time to initiate cooldown
]
j Cooldown Operator action time to initiate depressurization Depressurization Operator action time to initiate SI termination i
SI termination and pressure equalization l
l 90238:10/081985 10
WESTINGHOUSE CLASS 3 TABLE 11.2 GINNA SPECIFIC OPERATOR ACTION TIMES I-l l
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WESTINGHOUSE CLASS 3 D.
Transient Description - Case 1 S
The sequence of events for Case 1 is presented in Table 11.3.
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Following the tube rupture, reactor coolant flows from the primary into the secondary side of the ruptured steam generator since the primary j
pressure is greater than the steam generator pressure.
In response to this loss of reactor coolant, pressurizer level decreases as shown in j
Figure 11.4.
The RCS pressure also decreases as shown in Figure 11.1 as the steam bubble in the pressurizer expands.
As the RCS pressure decreases due to the continued primary to secondary leakage, automatic reactor trip occurs on a overtemperature delta-T setpoint.
.i Af ter reactor trip, core power rapidly decreases to decay heat levels.
The turbine stop valves close and steam flow to the turbine is l
terminated. The steam dump system is designed to actuate following i
reactor trip to limit the increase in secondary pressure, but the steam dump valves remain closed due to the loss of condenser vacuum resulting l
from the assumed loss of offsite power at the time of reactor trip.
- Thus, the energy transfer from the primary system causes the secondary side pressure to increase rapidly af ter reactor trip until the steam generator PORVs lift to dissipate the energy, as shown in Figure 11.3.
The RCS pressure decreases more rapidly af ter reactor trip as energy transfer to the secondary shrinks the reactor coolant and the leak flow continues to deplete primary inventory.
The decrease in RCS inventory results in a low pressurizer pressure SI signal.
Pressurizer level also decreases more rapidly following reactor trip and the pressurizer eventually empties as shown in Figure 11.4.
Af ter the pressurizer empties, the RCS pressure rapidly decreases as shown in Figure 11.1.
Since offsite power is assumed lost at reactor trip, the RCPs trip and a gradual transition to n,atural circulation flow occurs.
Immediately following reactor trip the temperature rise across the core decreases as 90238:lD/081985 12
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WESTINGHOUSE CLASS 3 core power decays (see Figure 11.2), however, the temperature rise subsequently increases as natural circulation flow develops.
The cold leg temperatures trend toward the steam generator temperature as the fluid 4
residence time in the tube regicn increases.
The RCS temperatures continue to slowly decrease due to the continued addition of the auxiliary feedwater to the steam generators until operator actions are initiated to I
cool down the RCS.
i Major Operator Actions 1.
Identify and Isolate the Ruptured Steam Generator Once a tube rupture has been identified, recovery actions begin by i
isolating steam flow f rom the ruptured steam generator and throttling i
l the auxiliary feedwater flow to the ruptured steam generator. As indicated previously the ruptured steam generator is assumed to be a,c identified and isolated when the narrow range level reaches ontherupturedsteamgeneratororat[}#minutesafter
~ '
i initiation of the SGTR, whichever is longer.
For Ginna, the time to a,c a,c reach [
is less than minutes, thus the ruptured steam generator is assumed to be isolated at minutes.
2.
Cool down the RCS to Establish Subcooling Margin
- a,c After isolation of the ruptured steam generator, there is a minute operator action time imposed prior to cooldown. The actual delay time used in the analysis is 4 seconds longer because of the computer program requirements for simulating the operator actions.
After this time, the RCS is cooled as rapidly as possible by dumping steam from the intact steam generators.
Since offsite power is lost, the RCS is cooled by dumping steam to the atmosphere using the PORV on
[
the intact steam generator.
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WESTINGHOUSE CLASS 3 The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20*F plus an allowance of 17'F for subcooling uncertainty.
3 This cooldown ensures that there will be
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adequate subcooling in the RCS after the subsequent depressurization of the RCS to the ruptured steam generator pressure. The reduction in the intact steam generator pressure required to accomplish the cooldown is shown in Figure 11.3, and the effect of the cooldown on the RCS temperature is shown in Figure 11.2.
The RCS pressure also decreases during this cooldown process due to shrinkage of the reactor coolant as shown in Figure 11.1.
3.
Depressurize RCS to Restore Inventory i
NO After the RCS cooldown, a [ ] minute operator action time is included j
prior to depressurization. The RCS is depressurized at 2542 seconds to assure adequate coolant inventory prior to terminating SI flow.
With the RCPs stopped, normal pressurizer spray is not available and j
thus the RCS is depressurized by opening a pressurizer PORV. The depressurization is continued until any of the following conditions 4
are satisfied: RCS pressure is less than the ruptured steam generator pressure and pressurizer level is greater than 0% plus an allowance of 35 for pressurizer level uncertainty, or pressurizer level is greater than 80% minus an allowance of 3% for pressurizer level uncertainty, or RCS subcooling is less than the 17*F allowance for subcooling uncertainty. The RCS depressurization reduces the break flow as shown in Figure II.6 and increases SI flow to refill the pressurizer, as i
shown in Figure 11.4.
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4.
Terminate SI to Stop Primary to Secondary Leakage 1
l The previous actions should have established adequate RCS subcooling, verified a secondary side heat sink, and restored the reactor coolant inventory following an SGTR to ensure that SI flow is no longer needed. When these actions have been completed, the SI flow must be l
l 90238:10/082085 14
T WESTINGHOUSE CLASS 3 1
stopped to prevent repressurization of the RCS and to terminate l
primary to secondary leakage. The SI flow is terminated when the RCS pressure increases, minimum AFW flow is available and at least one i
intact steam generator level is in the narrow range, RCS subcooling is greater than the 17'F allowance for subcooling uncertainty, and the pressurizer level is greater than the 3% allowance for pressurizer level uncertainty. To assure that the RCS pressure is increasirg, SI was not terminated in the analysis until the RCS pressure increased to 50 psi above the ruptured steam generator pressure.
aL Af ter depressurization is completed, an operator action time of [ ]
minute is imposed prior to SI termination.
After SI is terminated, i
break flow continues to accumulate in the secondary side resulting in
]
overfill of the ruptured steam generator at 3336 seconds.
For the period af ter overfill occurs, the amount of water released to the i
atmosphere via the ruptured steam generator PORV is considered equal I
to the break flow. The primary to secondary leakage continues af ter the SI flow is terminated until the RCS and ruptured steam generators equalize. This occurs when the intact steam generator PORV is locally opened to cooldown the RCS so that subcooling may be maintained. When the PORV is opened the increased energy transfer f rom primary to j
secondary depressurizes the RCS to the ruptured steam generator
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pressure.
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TABLE 11.3 SEQUENCE OF EVENTS CASE 1 1
EVENT Time (sec)
? f-Reactor Trip Ruptured SG Isolated Intact SG PORV Opened Intact SG PORY Isolated j
d i
PRZR PORY Opened i
PRZR PORV Closed i
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SI Terminated Overfill Ruptured SG Intact PORY Opened Break Flow Terminated l
90238:10/082085 16
WESTINGHOUSE CLASS 3 a,r
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9023B:10/082085 22
WESTINGHOUSE CLASS 3 E.
Transient Description - Case 2
[
]a,c Thus, the Case 2 transient is identical to the Case 1 transient until that time. The sequence of events for Case 2 is presented in Table 11.4.
Following the tube rupture RCS pressure decreases as shown in Figure 11.7 due to the primary to secondary leakage.
In response to this depressurization, the reactor trips on overtemperature delta-T. After a
reactor trip, core power rapidly decreases to decay heat levels and the RCS depressurization becomes more rapid. The steam dump system is inoperable due to the assumed loss of offsite power, which results in the secondary pressure rising to the steam generator PORV setpoint as shown in Figure 71.9.
The decreasing pressurizer pressure leads to an automatic SI signal on low pressurizer pressure.
Pressurizer level also decreases more rapidly following reactor trip until it eventually empties, as shown in Figure 11.10.
Major Operator Actions 1.
'.dentify and Isolate the Ruptured Steam Generator As with Case 1 the ruptured steam generator is assumed isolated at minutes.
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9023B:10/082185 23
WESTINGHOUSE CLASS 3 2.
Cool Down the RCS to establish Subcooling Margin s
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The depressurization of the ruptured steam generator affects the RCS cooldown target temperature since the temperature is dependent upon the pressure in the ruptured steam generator. Since of f site power is lost the RCS is cooled by dumping steam to the atmosphere using the intact steam generator PORV. The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20*F plus an allowance of 17*F for instrument uncertainty. Because of the lower pressure in the ruptured steam generator the associated temperature the RCS must be cooled to is also lower, which has the net effect of extending the time for cooldown.
For Case 2 cooldown begins at 1806 seconds and is completed at 2626 seconds.
The reduction in the intact steam generator pressure required to accomplish the cooldown is shown in Figure 11.9, and the effect of the cooldown on the RCS temperature is shown in Figure 11.8.
The RCS pressure also decreases during this cooldown process due to shrinkage of the reactor coolant as shown in Figure 11.7.
3.
Depressurize to Restore Inventory o.
Af ter the RCS cooldown, a [ ]'cminute operator action time is included prior to depressurization. The RCS is depressurized at 2748 seconds to assure adequate coolant inventory prior to terminating SI flow.
With the RCPs stopped, normal pressurizer spray is not available and thus the RCS is depressurized by opening a pressurizer PORV.
The depressurization is continued until any of the following conditions are satisfied: RCS pressure is less than the ruptured steam generator pressure and pressurizer level is greater than 0% plus an allowance of 3% for pressurizer level uncertainty, or pressurizer level is greater than 80% minus an allowance of 3% for pressurizer level uncertainty, or RCS subcooling is less than the 17*F allowance for subcooling i
9023B:lD/082085 24
WESTINGHOUSE CLASS 3 uncertainty. The RCS depressurization reduces the break flow as shown in Figure 11.12 and increases SI flow to refill the pressurizer, as shown in Figure 11.10.
4.
Terminate Si to Stop Primary to Secondary Leakage The previous actions should have established adequate RCS subcooling, verified a secondary side heat sink, and restored the reactor coolant inventory following an SGTR to ensure that SI flow is no longer needed. When these actions have been completed, the SI flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary leakage. The SI flow is terminated when the RCS pressure increases, minimum AFW flow is available and at least one intact steam generator level is in the narrow range, RCS subcooling is greater than the 17'F allowance for subcooling uncertainty, and the pressurizer level is greater than the 3% allowance for pressurizer level uncertainty. To assure that the RCS pressure is increasing, SI was not terminated until the RCS pressure increased to 50 psi above the ruptured steam generator pressure.
Af ter depressurization is completed, an operator action time of
[ ]#'C minute is imposed prior to SI termination.
Figure 11.12 shows that the primary to secondary leakage continues after the SI flow is stopped until the RCS and ruptured steam generator pressure equalize.
For Case 2, the ruptured steam generator does not overfill.
4 9023B:10/082085 25
WESTINGHOUSE CLASS 3 TABLE 11.4 SEQUENCE OF EVENTS CASE 2 EVENT TIME (sec) a f.
r-Reactor Trip Ruptured SG Isolated l
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l PRZR PORV Opened l
PRZR PORV Closed l
i SI Terminated i
Break Flow Terminated
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9023B:10/082085 26
WESTINGHOUSE CLASS 3 n.c 8
CORE PESSURE - CASE 2 FIGURE II.7 90238:10/082085 27
WESTINGHOUSE CLASS 3 3,c.
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'l MACT LOOP HOT A'O COLD LEG RCS TER.RATURES CASE 2 FIGURE II.8 i
90238:10/082085 28
WESTINGHOUSE CLASS 3 as UXPS A #O B SECGDR( PRESSUE - CASE 2 FIGURE II.9 90238:10/082085 29
WESTINGHOUSE CLASS 3 3.'
PRESSURIZERLEVEL-CASE 2 FIGURE II.10 90238:1D/082085 30
I WESTINGHOUSE CLASS 3 i
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PRil%RY TO SECGOARY ifXKE - CASE 2 FIGURE 11.12 90238:10/082005 32 w
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WESTINGHOUSE CLASS 3 F.
Mass Releases The mass releases were determined for each of the single failure cases for
{
use in evaluating the exclusion area boundary and low population zone radiation exposure. The steam releases from the ruptured and intact steam generators, the feedwater flows to the ruptured and intact steam generators, and primary to secondary break flow into the ruptured steam generator were determined for the period f rom accident initiation until 2 l
hours after the accident and from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident. The releases for 0-2 hours are used to calculate the radiation doses at the exclusion area boundary for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exposure, and the releases for 0-8 l
hours are used to calculate the radiation doses at the low population zone for the duration of the accident.
i In the LOFTTR1 analyses, the SGTR recovery actions in the E-3 guideline
)
were simulated until the termination of primary to secondary leakage.
After the primary to secondary leakage is terminated, the operators will continue the SGTR recovery actions in the E-3 guideline to prepare the i
plant for cooldown to cold shutdown conditions. These actions include establishing normal Chemical and Volume Control System (CVCS) operation to provide reactor coolant inventory control and a boration path; restarting a reactor coolant pump (RCP), if none are running, to ensure homogeneous I
RCS conditions and to provide normal pressurizer spray; or stopping one RCP, if both are running, to minimize the heat input during the subsequent I
cooldown; and the actions necessary to minimize the spread of contamination on the secondary side. When the instructions provided in E-3 are completed, the plant should be cooled and depressurized to cold l
shutdown conditions. There are three alternate means of performing the I
post-SGTR cooldown provided in the WOG Emergency Response Guidelines.
The guidelines are:
ES-3.1, POST-SGTR COOLDOWN USING BACKFILL; ES-3.2, 4
)
POST-SGTR COOLDOWN USING BLOWDOWN; and ES-3.3, POST-SGTR C00LDOWN USING STEAM 00MP. The preferred methods are using backfill or blowdown since these methods minimize the radioactivity released to the atmosphere.
The ES-3.3 guideline using steam dump provides the f astest method for j
depressurizing the RCS and ruptured steam generator.
This method also results in the worst radiological releases, especially if steam dump to i
90238:10/082085 33
WESTINGHOUSE CLASS 3 the condenser is unavailable. Therefore, the method using steam dump was selected for evaluation of the long-term mass releases since this produces conservative results for the of fsite dose evaluation.
It is noted that the use of the steam dump method would not be permitted if steam generator overfill occurs and water enters the main steamlines.
The high level actions for the ES-3.3 guideline are discussed below.
1.
Prepare for Cooldown to Cold Shutdown The initial steps to prepare for cooldown to cold shutdown are performed in the E-3 guideline following SI termination, and these steps will be continued in ES-3.3 if they have not already been I
completed. A few additional steps are also performed in ES-3.3 prior to initiating cooldown. These include isolating the cold leg 51 accumulators to prevent unnecessary injection, energizing pressurizer heaters as necessary to saturate the pressurizer water and to provide for better pressure control, and assuring adequate shutdown margin in the event of potential boron dilution due to in-leakage from the ruptured steam generator.
2.
Cooldown RCS to Residual Heat Removal (RI:R) System Temperature The RCS is cooled by steaming and feeding the intact steam generator similar to a normal cooldown. Since all immediate safety concerns have been resolved, the cooldown rate should be maintained less than the maximum allowable rate of 100*F/hr. The preferred means for cooling the RCS is steam dump to the condenser since this minimizes the radiological releases and conserves feedwater supply.
The PORV for the intact steam generator can also be used if steam dump to the condenser is unavailable. When the RHR system operating temperature is reached, the cooldown is stopped until RCS pressure can also be decreased. This ensures that pressure / temperature limits will not be exceeded.
90238:10/082085 34
WESTINGHOUSE CLASS 3 3.
Depressurize RCS to RHR System Pressure When the cooldown to RHR system temperature is completed, the pressure in the ruptured steam generator is decreased by releasing steam from the ruptured steam generator. Steam release to the condenser is preferred since this minimizes radiological releases. However, steam can also be released to the atmosphere using the PORV on the ruptured steam generator. An evaluation of the potential radiological consequences should be performed before releasing steam from the ruptured steam generator to the atmosphere. As the ruptured steam generator pressure is reduced, the RCS pressure is maintained equal to the pressure in the ruptured steam generator in order to prevent in-leakage of secondary side water or additional primary to secondary leakage. Normal pressurizer spray is the preferred means of RCS pressure control since this conserves coolant inventory.
If pressurizer spray is not available, a pressurizer PORV or auxiliary spray can be used to control RCS pressure.
When overfill of the ruptured steam generator occurs, as with Case 1, guideline ES-3.1 POST-SGTR C00LODWN USING BACKFILL is assumed to be used.
The high level actions for ES-3.1 are similar to ES-3.3.
However, the method by which ES-3.1 instructs the operator to depressurize the ruptured steam generator differs from ES-3.3.
In Guideline ES-3.1 the RCS is depressurized to promote back flow through the failed tube which depressurizes the ruptured steam generator without steam releases to the atmosphere.
4.
Cooldown to Cold Shutdown When RCS temperature and pressure have been reduced to the RHR system in-service values, RHR system cooling is initiated to complete the cooldown to cold shutdown. When cold shutdown conditions are achieved, the pressurizer can be cooled to terminate the event.
9023B:1D/082005 35
WESTINGHOUSE CLASS 3 F.1 Methodology For Calculation Of Mass Releases a/-
I l
t
?
I
\\
i I
l l
B l
90238:10/082085 36
~_.
WESTINGHOUSE CLASS 3 oc i
I 1
l
)
I l
t I
l l
i I
l i
I i
L 1
F.2 Mass Release Results 1
[
The mass release calculations were performed for both single failure cases using the methodology discussed above. For the time period from initiation of 4
the accident until leakage termination, the releases were determined from the l
4 LOFITRI results for two separate periods for use in the dose calculations.
The first time period considered is from accident initiation until reactor i
trip. Since the condenser is in service until reactor trip, any radioactivity l
90238:10/082185 37 i
WESTINGHOUSE CLASS 3 released to the atmosphere prior to reactor trip will be through the condenser air ejector. Af ter reactor trip, the releases to the atmosphere are assumed to be via the steam generator PORVs. The mass releases calculated from the time of leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and from 2-8 hours are also assumed to be released to the atmosphere via the steam generator PORVs. The mass releases for the SGTR event [
]a,c (Case 1) are presented in Table 11.5.
The results indicate that approximately [
]
lba of steam and [
] lbs of water is released from the ruptured steam generator to the atmosphere in the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. A total of [
] lbm of primary water is transferred to the secondary side of the ruptured steam generator before the break flow is terminated.
i I'
The mass releases for the SGTR event assuming
,s.;
i (Case 2) 1 are presented in Table 11.6.
The results indicate that approximately [
]
] lbe of steam is released to the atmosphere from the ruptured steam
]
generator within the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> [
]
lbm is 1
released to the atmosphere from the ruptured steam generator. A total of
[
] lbs of primary water is transferred to the secondary side of the ruptured steam generator before break flow is terminated.
I i
i i.
I i
i i
90238:10/082185 38
_ -. -.. - ~
~.
i j
WESTINGHOUSE CLASS 3 f
TABLE 11.5
)
CASE 1 MASS RELEASES TIME PERIOD 0-TRIP TRIP TMSEP -
OVFILL - TTBRK -
T2 HRS TMSEP OVFILL TTBRK T2 HRS 1RHR Faulted SG
._ c Condenser f
i Atmosphere 1
1 Feedwater f
i l
Intact SG l
Condenser Atmosphere Feedwater j
i Break Flow I
I PA i
TRIP
= Time of reactor trip = _ _sec.
_ p,c TMSEP = Time when water reaches the moisture separators =_
sec.
_y OVFILL = Time when steam generator overfills =
sec.
TTBRK = Time when break flow is terminated =-
-- <, 4 sec.
1 T2 HRS = Time at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> = 7200 sec.
~
~
TRHR
= ilme to reach RHR in-service conditions, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> = 28,800 sec.
4 l
l 90238:10/082005 39 t
WESilNGHOUSE CLASS 3 TABLE 11.6 CASE 2 MASS RELEASES TOTAL MASS FLOW (POUNDS)
TIME PERIOD 0-TRIP 1 RIP TMSEP -
ITBRK -
T2 HRS TMSEP TTBRK 12 HRS TRHR Faulted SG l
_* l-Condenser I
Atmosphere i
Intact SG l
i I
1 I
l i
Condenser
{
j Atmosphere Feedwater i
Break Flow l
.'h'-
TRIP
Time of reactor trip
sec.
4
-3 j.
TMSEP = Time when water reaches the moisture separators =
sec.
2
_ac TT8RK = Time when break flow is terminated =
sec.
T2 HRS = Time at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> - 7200 sec.
TRHR
= Time to reach RHR in-service conditions, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> - 28,800 sec.
90238:10/082005 40
WESTINGHOUSE CLASS 3 TABLE II.7 SUMMARIZED MASS RELEASES TOTAL MASS FLOW (POUNDS)
CASE 1 CASE 2 0-TTBRK -
2HR'S -
0-TTBRK -
2 HRS TTBRK 2 HRS 8 HRS TTBRK 2 HRS 8 HRS Faulted SG Condenser 47,800 0
0 47,800 0
0 l
Atmosphere 37,600 0
0 89,500 17,700 0
Feedwater 76,400 0
0 91,300 0
0 Intact SG Condenser 47,200 0
0 47,200 0
0 Atmosphere 79,200 148,800 566,400 67,600 174,200 444,500 Feedwater 160,900 155,200 575,000 152,400 179,100 444,500 Break Flow 121,000 0
0 166,400 0
0 9023B:10/082085 41
WESTINGHOUSE CLASS 3 Ill.
RADIOLOGICAL CONSEQUENCES ANALYSIS The evaluation of the radiological consequences of a steam generator tube rupture, assumes that the reactor has been operating at the proposed Technical Specification limit for primary coolant activity and at the existing Technical Specification limit for primary to secondary leakage for sufficient time to ostablish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant. Radionuclides f rom the primary coclant enter the steam generator, via the ruptured tube, and are released to the atmosphere through the steam generator safety or power operated relief valves and via the condenser air ejector exhaust.
i l
The quantity of radioactivity released to the environment, due to a SG1R, l
depends upon primary and secondary coolant activity, iodine spiking effects, primary to secondary break flow, break flow flashing fracticns, attenuation of iodine carried by the flashed portion of the break flow, partitioning of iodine between the liquid and steam phases, the mass of fluid released from the generator and liquid-vapor partitioning in the turbine condenser hot well. All of these parameters were conservatively evaluated for a design basis double ended rupture of a single tube.
A.
Desian Basis Analytical Assumptions The major assumptions and parameters used in the analysis are itemized in Table 111.1.
The following is a discussion of the source term.
Source Term Calculations The radionuclide concentrations in the primary and secondary system, prior to and following the SGTR are determined as follows:
a.
The iodine concentrations in the reactor coolant will be based upon preaccident and accident initiated iodine spikes.
90238:10/082085 42
j WESTINGHOUSE CLASS 3 I
j i.
Accident Initiated Spike - The initial primary coolant iodine concentration is 1 pCi/gm of Dose Equivalent (D.E.) 1-131.
]
Following the primary system depressurization associated with the SGTR, an iodine spike is initiated in the primary system which i
increases the iodine release rate f rom the fuel to the coolant to a I
value 500 times greater than the release rate corresponding to the initial primary system iodine concentration.
The duration of the spike,[
j,c hours, is sufficient to increase the initial RCS a
I-131 inventory by a factor of [
]a,c,
1 i
and has raised the primary coolant iodine concentration from 1 to 60 4
l pCi/ gram of D.E.1-131.
j b.
The initial secondary coolant iodine concentration is 0.1 pCi/ gram of I
D. E. 1-131.
i i
c.
The chemical form of iodine in the primary and secondary coolant is assumed to be elemental.
4 j
f Dose Calculations i
The following assumptions and parameters were used to calculate the activity released to the atmosphere and the offsite doses following a SGTR.
l 1.
The mass of reactor coolant discharged into the secondary system through the rupture and the mass of steam released f rom the intact and ruptured steam generators to the atmosphere are presented in Table 11.5 and 11.6.
1 l
2.
The time dependent fraction of rupture flow that flashes to steam and is immediately released to the environment is presented in Figure 111.1.
]
3.
The time dependent iodine removal ef ficiency for scrubbing of steam bubbles as they rise from the leak site [(
f
)] to the water surface was also determined for each case.
[
90238:10/082185 43
_ -. =. --
WESTINGHOUSE CLASS 3
}a,c The iodine removal efficiencies are shown in Figure 111.2.
4.
The 0.2 gpm primary to secondary leak is assumed to be split evenly between the steam generators.
5.
The iodine partition factor between the liquid and steam of the ruptured and intact steam generators is assumed to be 100.
6.
No credit was taken for radioactive decay during release and transport, or for cloud depletion by ground deposition during transport to the site boundary or outer boundary of the low population zone.
7.
Short-term atmospheric dispersion factors (x/Qs) for accident analysis and breathing rates are provided in Table 111.4. The breathing rates were
}
obtained from NRC Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for Pressurized Water l
Reactors", Rev. 2, June 1974.
i l
Offsite Thyroid Dose Calculation Model
[
Offsite thyroid doses are calculated using the equation:
f Y
~
g { (IAR)q (BR))
(x/Q)
D DCF
=
Th i
j
~
~
i where t
j (IAR)$)
integrated activity of isotope i released during the time interval j in Ci*
i l
No credit is taken for cloud depletion by ground deposition or by radioactive decay during transport to the exclusion area boundary or to j
the outer boundary of the low-population zone.
90238:10/082085 44 I
l
i WESTINGHOUSE CLASS 3 4
l (BR)3 breathing rate during time interval j in
=
3 meter /second (Table 111.4) 4 i
4 (x/Q))
atmospheric dispersion f actor during time interval j
=
j in second/ meter 3 (Table 111.4)
I j
(DCF)g thyroid dose conversion factor via inhalation for
=
i isotope i in res/Ci (Table 111.5) thyroid dose via inhalation in rem D
=
Th e
i Results
}
f Thyroid doses at the Exclusion Area Boundary and Low Population Zone are l
presented in Table 111.6. All doses are well within the guidelines of I
)
i 1
I i
4
}
2 i
i l
I i
4 1
l f
l I
90238:10/082085 45
WESTINGHOUSE CLASS 3 TABLE 111.1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE 1.
Source Data A.
Core power level, MWt 1520 B.
Total steam generator tube 0.2 leakage, prior to accident, gpm C.
Reactor coolant iodine activity:
1.
Accident Initiated Spike The initial RC iodine activities based on 1 pCi/ gram of 0.E. 1-131 are presented in l
Table 111.3.
The iodine appearance rates assumed for the accident initiated spike are presented in Table 111.2.
2.
Pre-Accident Spike Primary coolant iodine activities based on 60 pCi/ gram of D.E.1-131 are presented in Table 111.3.
D.
Secondary system initial activity Dose equivalent of 0.1 pCi/gm of I-131, presented in Table !!!.3.
8 E.
Reactor coolant mass, grams 1.27 x 10 90238:10/082085 46
WESTINGHOUSE CLASS 3 TABLE 111.1 (Sheet 2)
I F.
Steam generator mass 3.39 x 10 I
(each), grams 6.
Offsite power Lost at time of reactor
]
trip l
i H.
Primary-to-secondary leakage 8
duration for intact SG, hrs.
Y I
1.
Species of iodine 100 percent elemental i
l II.
Activity Release Data i
j A.
Faulted steam generator i
i 1.
Rupture flow See Table 11.5 or II.6 i
1 i
2.
Rupture flow flashing fraction See Figure 111.1 1
1 i
l 3.
Iodine scrubbing plus moisture See Figure III.2
[
separator removal ef ficiency
}
4.
Total steam release, lbs See Table 11.5 or II.6 5.
Iodine partition factor I
a.
Prior to overfill 100 t
i b.
After overfill 1.0 - See Figure 111.3
(
6.
Location of tube rupture
[
]a,c l
t I
l l
90238:10/082085 47
-#4.w-,..--,,re--e,m-+
w re
,w.-.w--
r--s-
-+--m3.,ww----
,e,*
,,ea---.
,e--_eee-,
--r,-m-
-.ww r-e-.-
-w.,,vm,--,,v--,
WESTINGHOUSE CLASS 3 TABLE III.1 (Sheet 3)
B.
Intact steam generator 1.
Primary-to-secondary leakage, gpm 0.1 2.
Total steam release, lbs See Table 11.5 or II.6 3.
Iodine partition factor 100 C.
Condenser l.
Iodine partition factor 100 D.
Atmospheric Dispersion Factors See Table 111.4 i
i 4
i j
i i
e i
i s
1 90238:10/082085 48 1
- c.,. 1 WES11NGH00SE CLASS 3
<e.
O TABLE III.2 IODINE SPIKE APPEARANCE RATES (CURIES /SECONO) 4
,i 1
"k-131
-I-132 1-133 1-134 I-135
? -
1 l
'O.94 2.22 1.74 3.07 2.34 0
l 1
t
(
i i
i
+
4 i
i i
i i
1 l
l 1
is 9023B:10/082085 49
~ - - -.. -...
WE511NGHOUSE CLASS 3 TABLE 111.3 IODINE SPECIFIC ACTIVITIES IN (pCi/gm) THE PRIMARY AND SECONDARY COOLANT BASED ON 1, 60 AND 0.1 pCi/ gram 0F D.E.1-131 Primary Coolant Secondary Coolant Nuclide 1 uCi/cm 60 pCi/cm 0.1 uCi/gm 1-131 0.79 47.1 0.079
(
I-132 0.35 20.7 0.035 I-133 1.01 60.7 0.101 1-134 0.20 12.2 0.020 I-135 0.79 47.1 0.079 a
1 9023B:10/082085 50
WESTINGHOUSE CLASS 3 TABLE 111.4 ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES Time Exclusion Area Boundary Low Population Breathing (hours) x/Q (Sec/m )
Zone x/Q (Sec/m )
Rate (m /Sec) [4]
-4
-4 0-2 4.8 x 10 3 x 10~
3.47 x 10
-5
-4 2-8 3 x 10 3.47 x 10 1
9023B:10/082085 51
WESTINGHOUSE CLASS 3 1ABLE 111.5 THYROID DOSE CONVERSION FACTORS (Rem / Curie) [5]
Nuclide 6
I-131 1.49 x 10 4
I-132 1.43 x 10 1-133 2.69 x 10 l
3 I-134 3.73 x 10 4
1-135 5.60 x 10 I
s i
9023B:10/082085 52
1 1
l WESTINGHOUSE CLASS 3 TABLE 111.6 RF.SULIS
\\
\\
Doses (Rem)
\\
Case 1 Case 2 1.
Accident Initiated Iodine Spike Exclusion Area Boundary (0-2 hr.)
n.c Thyroid 4.9 Low Population Zone (0-8 hr.)
Thyroid 0.3 2.
Pre-Accident Iodine Spike Exclusion Area Boundary (0-2 hr.)
Thyroid 30.3 Low Population Zone (0-8 hr.)
Thyroid 1.9 9023B:10/082085 53
WESTINGHOUSE CLASS 3 c
b RE FW FLA900 FET!:h - CE I a,c e
O FIGURE III.1 54
WESTINGHOUSE CLASS 3 a.C 4
G e
4 4
~
_ Q,C o
SCRUBBING AND SEPARATER REMOVAL EFFICIENCY FIGURE III.2 55
WESTINGHOUSE CLASS 3
_ a.c.
G m
m B
I m
FAULTED SG IODINE PARTITION FACTORS 4
FIGURE III.3 56
IV.
CONCLUSION The potent'ial radiological consequences of a steam generator tube f ailure were evaluated for the R.E. Ginna nuclear power plant to demonstrate that the use of the Standard Technical Specification (STS) primary coolant activity limit of 1 pCi/ gram of dose equivalent 1-131 will result in of fsite doses that are within the appropriate guidelines. The mass releases for a design basis double ended rupture of a single tube with a loss of offsite power were conservatively calculated using the computer code LOFTTRI.
Two chses were considered:
ts-The analysis explicitly modeled the time needed for the operators to perform the recovery steps outlined in guideline E-3 of Revision 1 of the Westinghouse Owners Group Emergency Response Guidelines. The resulting doses at the exclusion area boundary and low population zone are within the allowable guidelines as specified by Standard Review Plan 15.6.3 and 10CFR100. Consequently, the STS primary coolant activity limit is sufficiently low to ensure that the radiological consequences of a steam generator tube rupture at the R.E. Ginna plant will be within the guidelines.
9023B:10/082185 57
WESTINGHOUSE CLASS 3 V.
REFERENCES 1.
Lewis, Volpehein, Huang, Behnke, Fittante, Gelman, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"
WCAP-10750 and WCAP-10698, December 1984.
2.
Lewis, Huang, Rubin, " Evaluation of Offsite Radiation Doses for a Steam Generator Tube Rupture Accident," Supplement 1 to WCAP-10750 and WCAP-10698, May 1985.
3.
R. Elaisz, Letter f rom RGE to Westinghouse concerning Ginna specific operator action times for SGTR analysis, February 7,1985.
4.
NRC Regulatory Guide 1.4, Rev. 2, " Assumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for Pressurized Water Reactors", June 1974.
5.
NRC Regulatory Guide 1.109, Rev.1, " Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix I", October 1977.
(
90238:10/082185 58
_ _ _ _ _ _ _ _