ML20135A691
| ML20135A691 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/02/1996 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| VPNPD-96-103, NUDOCS 9612040005 | |
| Download: ML20135A691 (23) | |
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Wisconsin i Electnc POWER COMPANY 231 W MicNgort PO Box 2046, MNoukee. WI S32012046 (4141221 2345 VPNPD-96-103 10 CFR 50.4 10 CFR 50.90 December 2,1996 Document Control Desk US NUCLEAR REGULATORY COMMISSION l i
Mail Station PI 137 Washington, DC 20555-0001 i
Gentlemen: j 1
DOCKETS 50-266 AND 50-301 j EUPPLEMENT TO TECHNICAL SPECIFICATIONS j CIIANGE REOUESTS 188 AND 189 POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 This letter provides additional information in support of Technical Specifications Change Requests (TSCRs) 188 and 189. ,
TSCRs 188 and 189 were submitted in letters dated June 4,1996. Supplements to TSCRs have been submitted in letters dated l August 5,1996; September 26,1996; October 21,1996; November 13,1996; and November 20,1996. These requests I propose amendments to the Point Beach Technical Specifications that were identified by analyses performed in support of l Unit 2 operations following replacement of steam generators this fall. !
l We are providing additional information regarding the radiological consequences for the Steam Generator Tube Rupture, i Rupture of a Steam Pipe, Locked Rotor, and Rod Ejection accidents as attachments to this letter. This information is in response to your request for additional information dated November 13,1996.
We have determined that the additional information does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of any efiluent release, or result in any significant increase in individual or cumulative occupational exposure. Therefore, we conclude that the proposed amendments meet the requirements of 10 CFR S t.22(c)(9) and that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared. The original "No Significant Hazards" determinations for operation under the proposed Technical Specifications remain applicable.
If you require additional information, please contact us. \ g Sincerely, gg Bob _ nk Subscribjd and sprn before me on Vice Presi ent this 2n day of Llwd.o 1996.
Nuclear Power CAC 040005 cc: NRC Resident Inspector m <J 6fft NRC Regional Administrator r[Public, StaIe of Wisconsin I
9612040005 961202 My commission expires 80 /n / 2t,00 PDR ADOCK 05000266 # / ~
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATIONS CHANGE REQUESTS 188 AND 189 A request for additional information for Technical Specification Change requests 188 and 189 was transmitted to Wisconsin Electric in a letter dated November 13,1996. Part B of the request states the following:
Analyze the following accidents: ' (1) steam generator tube rupture, (2) control rod ejection, (3) locked rotor, and (4) main steamline break. Submit a copy of your accident analysis report that contains the calculated control room, EAB and LPZ doses and a comparison of the results tot he 10 CFR S0 and 100 :-3 -p=e criteria. Also provide sufficient documentation of your analyses to support an independent evaluation, such as code inputs, the references for code inputs (e.g. UFSAR or new docketed analyses), and the dose calculations.
The requested information is provided as follows:
Radiolonical Consecuences of a Locked Rotor Accident Introduction An instantanecus seizure of a reactor coolant pump rotor is assumed to occur which rapidly reduces flow through the afected reactor coolant loop. Fuel clad damage is assumed to occur as a result of this acculent. Due to the pressure differential between the primary and secondary systems, and assumed SG
- tube leaks, fission products are discharged from the primary into the secondary system. A portion of this radmactivity is released to the outside atmosphere through either the atmospheric dump valves (ADV) or safety valves (MSSVs). In addition, some of the iodine activity contained in the secondary coolant prior to the accident is released to atmosphere as a result of steaming of the SGs following the accident. This section describes the assumptions and analyses performed to determine the amount of radioactivity released and the oEsite and control room doses resulting from this release, laput Parameters and Assumptions The analysis of the locked rotor event radiological consequences uses the analytical methods and assumptions outlined in the Standard Resiew Plan (Reference 1).
One hundred and two (102) percent of the uprated power level of 1650 MWt (1683 MWt) is used in the analysis. For the pre-accident iodine spike it is assumed that a reactor transient has occurred prior to the locked rotor and has raised the RCS iodine concentration to 60 Ci/gm of dose equivalent (DE) 1-131.
Since fuel failure is assumed for this accident, it is not necessary to also assume an accident initiated spike, as is the case for events without fuel failure such as a SGTR or a MSLB. The noble gas activity concentration in the RCS at the time the accident occurs is based on a fuel defect level of 1.0% This is approximately equal to the Technical Specification value of 100/E bar pCi/gm fos. gross radioactivity.
The iodine activity concentration of the secondary coolant at the time the locked rotor occurs is assumed
- to be equivalent to the Technical Specification limit of 1.2 pCi/gm of DE I-131.
In deternuning the ossite and control room doses following the locked rotor, it is conservatively assumed that 100% of the fuel rods in the core suffer sufficient damage that all of their gap activity is released to the RCS. Ten percent of the total core activity for both iodines and noble gases is assumed to be in the fuel-cladding gap (Reference 2).
The total primary to secondary SG tube leak rate used in the analysis is the Technical Specification limit of 0.35 gpm per steam generator or 0,70 gpm total. No credit for iodine removal is taken for any steam released to the condenser prior to reactor trip and concurrent loss of offsite power. An iodine partition 1
I factor in the SGs of 0.01 (curies I /gm steam) / (curies I /gm water) is used (Reference 3). All noble gas l activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere At 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident the RHR System is assumed to be placed into service for heat removal, and
- there are no further steam releases to atmosphere from the secondary system.
The thyroid dose conversion factors, breathmg rates, and smospheric dispersion factors used in the dose calenlanaris are given in Table 1. The core and coolant actisities used in the radiological calculations are given in Table 2. The parameters associated with the control room HVAC modes are summanzed in Table 3. The remaining mapr assumptions and parameters used specifically in the locked rotor analysis are immind in Table 4.
Control Room Model The Point Beach control room HVAC system operates in one of four modes Mode 1 is the normal HVAC mode, in which 5% of the air flow is outside air and 95% is recirculated air. Mode 2, which consists of 100% recirculated unfiltered air within the control room, is initiated either by a containment isolation signal or manually from the control room. Mode 3 is initiated manually by operator action and allows for filtered recirculated airflow. Mode 4 is initiated either by a control room radioactisity signal or manually by operator action. In this mode,25% of the available flow is made up with filtered outside air while the remaining 75% air flow is unfiltered recirculation. The pirameters associated with the control room HVAC modes are summarized in Table 3. These parameters have been taken from Reference 4. In addition, a factor of 10 reduction to the thyroid dose is allowed with the use of Potassium lodide pills by
. the control room operators.
For the locked rotor accident it is assumed that the HVAC system begins in Mode 1. A containment isolation signal is never received so Mode 2 is not modeled for this accident scenario. The dose rates in the control room trip the control room monitors within 30 minutes, switching the system to Mode 4 where it remains throughout the event. The control room doses are calculated over a period of twenty-four hours to ensure that the largest doses to the control room operator are calculated since the ventilation system will continue to operate in the specified mode for several hours following the termination of the steam .
releases.
Description of Analyses The analysis of the locked rotor event radiological consequences uses the analytical methods and assumptions outlined in the Standard Review Plan (Reference 1). Because fuel failure is assumed, only a pre-accident iodine spike is assumed, rather than both pre-accident and accident initiated spikes, as is the
. case for events without fuel failure. !
Acceptance Criteria The dose limits for a locked rotor are a "small fraction of" the 10 CFR 100 guidelines values. A "small fraction of" is considered 10% of 10 CFR 100 guideline values, or 30 rem th 3Toid and 2.5 rem y-body.
The criteria defined in $RP Section 6.4 (Reference 5) are used for the control room dose limits: 30 rem l thvroid,5 rem whole body and 30 rem beta skin. Results The offsite and control room th>Toid, y-body, !
and beta skin ooses due to the locked rotor event are given in Table 5. I l
Comelusions l
7 The offsite thyroid and gamma body doses due to the locked rotor accident are within the acceptance
[ criteria. The control room gamma body and beta skin doses are within the current NRC acceptance !
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criteria for the control room. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> control room thyroid dose exceeds the 30 rem limit; however, assuming that the operators would be instructed to take the potassium iodide pills, the control room !
thyroid dose would be reduced to approximately 6.5 rem which is within the 30 rem limit.
i Areas of Conservatism with respect to the Steam Generator Replacement Project This analysis was performed to support both the Steam Generator Replacement Project and the fuel ,
upgrade /uprating program for Point Beach Units I and 2. As such, there are several areas in which a i more conservative or bounding value was used to support the fuel upgrade or uprating which would not be I necessary to support the replacement steam generators alone. This section describes the conservatisms i-ir-.;d into the locked rotor accident radiological analysis with respect to the Steam Generator i P-y= =t Project. <
)'
The steam releases for the locked rotor accident were calculated using the increased power level of 1650 MWt, a higher Tavg of $80*F, and an RCS pressure of 2250 psi. The steam releases calculated with these parameters bound the steam releases which would correspond to a power level of 1520 MWt, a Tavg of 573.9'F and an RCS pressure of 2000 or 2250 psi. Further, the source term calculations were performed i to incorporate the increased core thermal power level and fuel upgrade parameters. The upgraded fuel .
includes an increase in the mass of the fuel and enrichment which results in an incree to several isotopes in the core and coolant activities. The number of fuel rods assumed to suffer sufficient danunge to release all of their gap fraction was increased from 86% to 100%. For the current fuel cycle, only 86% of rods are calculated to enter DNB. Additionally, the locked rotor accident was analyzed using the analytical methods and assumptions outlined in the Standard Resiew Plan (Reference 1). Specifically, this means the accident also incorporated a pre-accident iodine spike equivalent to 60 pCi/gm of DE I 131 in the coolant actisities.
References
- 1. NUREG-0800, Standard Review Plan 15.3.3,15.3.4, Revision 2, " Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break", July 1981.
- 2. US AEC Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors", May 1974.
- 3. NUREG-0800, Standard Review Plan 15.6.3, " Radiological Consequences of a Steam Generator Tube Rupture (PWR)", Rev. 2, July 1981.
- 4. Wisconsin Electric letter to NRC, VPNPD-96-099, " Supplement to Technical Specifications Change Requests 188 and 189 Point Beach Nuclear Plant Units 1 and 2," Bob Link, November 20,1996.
- 5. NRC SRP Section 6.4, " Control Room Habitability System", Rev 2 July 1981,"NUREG-0800.
Radiolonical Conseevences of a Red Election Accident Introduction I 1
It is assumed that a mechanical failure of a control rod mechanism pressure housing has occurred, resulting in the ejection of a rod cluster control assembly and drive shaft. As a result of the accident fuel clad damage and a small amount of fuel melt are assumed to occur. Due to the pressure differential between the primary and secondary systems, radioactive reactor coolant is discharged from the primary into the secondary system. A portion of this radioactisity is released to the outside atmosphere through either the main condenser or the atmospheric dump valves (ADV)/ safety valves (MSSVs). Some of the 3
Y-y -.. i iodme activity contained in the secondary coolant prior to the accident is released to atmosphere as a result of steaming of the SGs following the accident. Additionally, radioactive reactor coolant is
- discharged to the containment via the spill from the reactor vessel head. A portion of this radioactivity is !
released through containment leakage to the environment. This section describes the assumptions and analyses performed to determine the amount of radioactivity released and the offsite and control room
- doses resulting from these releases. .
laput Parameters and Assumptions i t
The effsite and control room doses following a rod ejection accident are determined using present-day NRC regulatorf requirements. This includes taking into account a pre-accident iodine spike. For the pre-accident iodine spike it is assumed that a reactor transient has occurred prior to the rod ejection and has raised the RCS iodine conantration to 60 pCi/gm of dose equivalent (DE) 1 131. The noble gas actisity raarensration in the RCS at the time the accident occurs is based on a fuel defect level of 1.0%. This is approximately equal to the Technical Specification value of 100/E bar pCi/gm for gross radioactivity. i The iodine activity concentration of the secondary coolant at the time the rod ejection accident occurs is !
assumed to be equivalent to the Technical Specification limit of 1.2 pCi/gm of DE I 131. !
l As a result of the rod ejection accident less than 10% of the fuel rods in the core undergo cladding damage. In determining the offsite and control room doses following rod ejection accident, it is conservatively assumed that 10% of the fuel rods in the core suffer sufficient damage that all of their gap activity is released to the RCS. Ten percent of the total core actisity for both iodines and noble gases is assumed to be in the fuel-cladding gap (Reference 1). '
A small fraction (i.e.,0.25%) of the fuel in the core is assumed to melt as a result of the rod ejection accident. One-half of the iodine activity in the melted fuel is released to the RCS, while all of the noble gas activity in the melted fuel is released to the RCS.
Conservatively, all the iodine and noble gas activity (from prior to the accident an'd resulting from the accident) is assumed to be in the RCS when determining offsite and control room doses due to the primary to secondary SG tube leakage, and all of the iodine and noble gas actisity is assumed to be in the containment when determining offsite and control room doses due to containment leakage. However, 50% of the iodine activity released to the containment is assumed to instantaneously plate out on l containment surfaces.
The total primary to secondary SG tube leak rate used in the analysis is the Technical Specification limit of 0.35 gpm per steam generator or 0.70 gpm total. Primary and secondary system pressure are equalized after 1500 seconds, thus terminating primary to secondary leakage in the SGs. No credit for iodine removal is taken for any steam released to the condenser prior to reactor trip and concurrent loss of offsite power. An iodine partition factor in the SGs of 0.01 curies /gm steam / curies /gm water is used (Reference 2). All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.
Steam release of the initial secondary coolant activity from the SGs following the rod ejection accident is based on the maximum relief rate of 6.664E6 lb/hr through the main steam safety valves and a steam release duration of 86 seconds. This results in a steam release of 158,200 lb.
The Technical Specification design basis containment leak rate of 0.4% by weight of containment air is used for the initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Thereafter the containment leak rate is assumed to be one-half the design i value, or 0.2%/ day (Reference 1). The thyroid dose conversion factors, breathing rates, and atmospheric l dispersion factors used in the dose calculations are given in Table 1. The core an5 coolant actisities used in the dose calculations we given in Table 2. The parameters associated with the control room HVAC 4
- v. .:
l 1: . .
I modes are summarized in Table 3. The remaining major assumptions and parameters used specifically in i
the rod ejection analysis are itemized in Table 6.
. Castrol Rooms Model The Point Beach control room HVAC system operates in one of four modes Mode 1 is the normal HVAC l- mode, in which 5% of the air flow is outside air and 95% is recirculated air. Mode 2, which consists of 100% recirculated unfiltered air within the control room, is initiated either by a containment isolation signal or manually from the control room. Mode 3 is initiated manually by operator action and allows for -
Altered recirculated airflow, Mode 4 is initiated either by a control room radioactisity signal or manually by operator action. In this mode,25% of the available flow is made up with filtered outside air while the
- remaining 75% air flow is unfiltered recirculation. The parameters associated with the control room HVAC modes are summarized in Table 3. These parameters have been taken from Reference 3. -In addition, a factor of 10 reduction to the thyroid dose is allowed with the use of Potassium Iodide pills by the control room operators. 1 For the rod ejection accident it is assumed that the HVAC system begins in Mode.l. On containment isolation, the system is automatically shiAed to Mode 2 which would occur within 5 minutes of event initiation. When the dose rates in the control room exceed the high radiation alarm setpoint, the system is automatically shiAed to Mode 4 where it remains throughout the event. The switch to Mode 4 will occur within 30 minutes. For simplicity, the analysis models a shift from Mode 1 to Mode 4 at 30 minutes without switching to Mode 2.
Description of Analyses Performed The analysis of the rod ejection event radiological consequences uses the analytical methods and '
assumptions outlined in Regulatory Guide 1,77 (Reference 1). Because fuel failure is assumed, only a pre-accident iodine spike is assumed, rather than both pre-accident and accident initiated spikes, as is the case for events without fuel failure.
Acceptance Criteria 3 i
The offsite dose limits for a rod ejection accident are "well within" the 10 CFR 100 guideline values, or 75 j rem thyroid and 6 rem whole body (Reference 4). The criteria defined in SRP Section 6.4 (Reference 5) will be used for the control room dose limits: 30 rem thyroid,5 rem whole body and 30 rem beta skin.
1 Results The offsite and control room thyroid, y-budy, and beta skin doses due to the rod e'jection accident are given in Table 7.
Conclusions The offsite thyroid and whole body doses and the control room whole body and beta skin doses are within the current NRC -~je- criteria for a steamline break accident. The control room thyroid dose exceeds the 30 rem limit; however, assuming that the operators would be instructed to take the potassium 1 modide pills, the control room thyroid dose would be reduced to approximately 12 rem which is within the ]
30 rem limit.
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-o 1 j3 .
1 i-Areas of Conservatism with respect to the Steam Generator Replacement Project
, This analysis was performed to support both the Steam Generator Replacement Project and the fuel upgrade /uprating program for Point Beach Units 1 and 2. As such, there are several areas in which a L more conservative or bounding value was used to support the fuel upgrade or uprating which would not be necessary to support the rd-* steam generators alone. This section describes the conservatisms incorporated into the rod ejection accident radiological analysis with respect to the Steam Generator na ==t Project.
The source term calculations were performed to incorporate the increased core thermal power level and fuel upgrade p ..;;ers. The upgraded fuel includes an increase in the mass of the fuel and enrichment which results in an increase to many isotopes in the core and coolant activities. Additionally, the rod ejection accident was analyzed using the analytical methods and assumptions outlined in Reg Guide 1.77 (Reference 1). Specifically, this means the accident also incorporated a pre-accident iodine spike equivalent to 60 pCi/gm of DE I-131 in the coolant activities.
References
- 1. US AEC Regulatory Guide 1.77," Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors", May 1974.
- 2. NUREG-0800, Standard Review Plan 15.6.3, " Radiological Consequences of a Steam Generator Tube Rupture (PWR)", Rev. 2, July 1981.
- 3. Wisconsin Electric letter to NRC, VPNPD 96-099, " Supplement to Technical Specifications Change Requests 188 and 189 Point Beach Nuclear Plant Units 1 and 2," Bob Link, November 20,19%. l
- 4. NUREG-0800, Standard Review Plan 15.4.8, Revision 2, " Spectrum of Rod Ejection Accidents (PWR)", July 1981. .
- 5. NRC SRP Section 6.4, " Control Room Habitability System", Rev 2, July 1981, NUREG-0800.
Radiolonical Conseamences of a Steam Generator Tube Ruoture Accident Introduction For the SGTR event, the complete severance of a single steam generator tube is assumed to occur. Due to the pressure differential between the primary and secondary systems, radioactive reactor coolant is discharged from the primary into the secondary system. A portion of this radioactisity is released to the outside atmosphere through either the main condenser, the atmospheric dump valves (ADV) or safety valves (MSSVs). In addition, some of the iodine activity contained in the secondary coolant prior to the accident is released to atmosphere as a result of steaming of the SGs following the accident. This section describes the assumptions and analyses performed to determine the amount of radioactisity released and the offsite and control room doses resulting from this release.
Input Parameters and Assumptions The analysis of the steam generator tube rupture (SGTR) radiological consequences uses the analytical methods and assumptions outlined in the Standard Resiew Plan (Reference 1).
One hundred and two (102) percent of the uprated power level of 1650 MWt (1683 MWt) is used in the J analysis. For the pre-accident iodine spike it is assumed that a reactor transient has occurred prior to the 6 !
O SGTR and has raised the RCS iodine concentration to 60 pCi/gm of dose equivalent (DE) 1-131. For the accident initiated iodine spike, the reactor trip associated with the SG'IR creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to the maximum equilibrium RCS Technical Specification concentration of 1.0 Ci/gm of DE I 131. The duration of the accident initiated iodine spike is 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The noble gas activity concentration in the RCS at the time the accident occurs is based on a fuel defect level of 1.0%. This is approximately equal to the Technical Specification value of 100/E bar pCi/gm for gross radioactivity. The iodine activity concentration of the secondary coolant at the time the SGTR occurs is assumed to be equivalent to the Technical Specification limit of 1.2 pCi/gm of DE l-131. The amount of primary to secondary SG tube leakage in the intact SG is assumed to be equal to the Technical Specification limit for a single SG of 0.35 spm. No credit for iodine removal is taken for any steam released to the condenser prior to reactor trip and concurrent loss of offsite power. An iodine partition factor in the SGs of 0.01 (curies 1/gm steam) / (curies 1/gm water) is used (Reference 1). All noble gas activity carried over to the secondary side through SG tube 'eakage is assumed to be immediately released to the outside atmosphere.
~
Flow through the ruptured SG tube is assumed to be terminated at 30 minutes following accident initiation due to operator action. Eight hours aAcr the accident the RHR System is assumed to be placed into service for heat removal, and there are no further steam releases to atmosphere from the secondary system.
The thyroid dose conversion factors, breathing rates, and atmospheric dispersion factors used in the dose calculations are given in Table 1. The core and coolant activities used in the dose calculations are given in Table 2. The parameters associated with the control room HVAC modes are summarized in Table 3.
The remaining major assumptions and parameters used specifically in this analysis are itemized in Table 8.
Control Room Model The Point Beach control room HVAC system operates in one of four modes. Mode 1 is the normal HVAC mode, in which 5% of the air flow is outside air and 95% is recirculated air. Mode 2, which consists of 100% recirculated unfiltered air within the control room, is initiated either by a containment isolation signal or manually from the control room. Mode 3 is initiated manually by operator action and allows for filtered recirculated airflow. Mode 4 is initiated either by a control room radioactisity signal or manually by operator action, la this mode,25% of the available flow is made up with filtered outside air while the
- remaining 75% air flow is unfiltered recirculation. The parameters associated with the control room HVAC modes are summarized in Table 3; These parameters have been taken from Reference 2. In addition, a factor of 10 reduction to the thyroid dose is allowed with the use of Potassium lodide pills by the control room operators.
For the steam generator tube rupture accident it is assumed that the HVAC system begins in Mode 1. On containment isolation, which is conservatively assumed to begin 10 minutes aAer event initiation. the system is automatically shiAed to Mode 2. When the dose rates in the control room exceed one of the radiation alarm setpoints, the system is automatically shiAed to Mode 4 and remains there until the radiation release has ended. The shiA to Mode 4 will occur within 30 minutes. In order to bound the total dose received by the operators, the radiological consequences were calculated for both Mode 2 and Mode 4 separately. The first case has a shift from Mode I to Mode 2 aAer 10 minutes, and remaining in Mode 2 throughout the event. The second case has a shiA from Mode I to Mode 4 aAer 30 minutes, then )
remaining in Mode 4 for the rest of the event. The control room doses are calculated over a period of twenty-four hours to ensure that the largest doses to the control room operator are calculated since the ,
ventilation system will continue to operate in the specified modes for several hours following the j termination of the steam releases.
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.s d Description of Analyses The analysis of the steam generator tube rupture (SGTR) radiological consequences uses the analytical methods and assumptions outlined in the Standard Review Plan (Reference 1) to calculate the actisity release to atmosphere from the ruptured and intact SGs and the resulting offsite apd control room doses.
Both the pre-accident iodine spike and acculent initiated iodine spike models are analyzed for these release paths.
Acceptance Criteria The offsite dose limits for a SGTR with a pre-accident iodine spike are the guideline values of 10 CFR 100 (Reference 1). These guideline values are 300 rem thyroid and 25 rem y-body. For a SGTR with an accident initiated iodine spike, the &= criteria are a "small fraction of" the 10 CFR 100 guideline values, or 30 rem thyroid and 2.5 rem y-body. The criteria defined in SRP Section 6.4 (Reference 3) will be used for the control room dose limits: 30 rem thyroid, 5 rem whole body and 30 rem beta skin.
Results The offsite and control room thyroid, y-body, and beta skin doses due to the SGTR accident are given in Table 9. The results of both control room models are included.
- Conclusions The offsite end control room doses due to the SGTR are within the acceptance criteria.
Antas of Conservatism with respect to the Steam Generator Replacement Pmject This analysis was performed to support both the Steam Generator Replacement Pioject and the fuel upgrade /uprating program for Point Beach Units I and 2. A similar analysis was also performed to support the Steam Generator Replacement Project; however, since additional calculations are necessary to address control room habitability issues the fuel upgrade /uprate SGTR analysis which is the most recent analysis is being used to support the Steam Generator Replacement Project at this time. As such, there are several areas in which a more conservative or bounding value was used to support the fuel upgrade or uprating which would not be necessary to support the replacement steam generators alone. This section describes the conservatisms incorporated into the steam generator tube rupture accident radiological I analysis with respect to the Steam Generator Replacement Project.
The thermal and hydraulic SGTR calculations were performed for both the replacement steam generator and the fuel upgrade /uprate programs. The primary to secondary break flow and the steam releases to the atmosphere for the fuel upgrade uprate are approximately 7% higher than those calculated for the Steam ;
Generator Replacement Project. These higher break flow and steam releases were used to calculate the j offsite and control room doses for the fuel upgrade /uprate program. In addition, the source term calculations were performed to incorporate the increased core thermal power level and fuel upgrade parameters. The upgraded fuel includes an increase in the mass of the fuel and enrichment which results in an increase to many isotopes in the core and coolant activities. Additionally, the steam generator tube
^
rupture accident was analyzed using the analytical methods and assumptions outlined in the Standard Review Plan (Reference 1). Specifically, this means the accident incorporated both an accident initiated and a pre-accident iodine spike in the coolant actitties; however, as consistent with the Standard Review Plan, the steam generator partition coefficient was reduced from 0.1 presented in the Point Beach FSAR to 0.01. -
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V-l References
- 1. NUREG 0800, Standard Review Plan 15.6.3, " Radiological Consequences of a Steam Generator Tube Rupture (PWR)", Rev. 2, July 1981.
- 2. Wisconsin Electric letter to NRC, VPNPD-96 099, " Supplement to Technical Specifications Change Requests 188 and 189 Point Beach Nuclear Plant Units I and 2," Bob Link, November 20,1996.
l
- 3. NRC SRP Section 6.4, " Control Room Habitability System", Rev 2, July 1981, NUREG-0800.
Radiolonical Consecuences of a Steamline Break Accident Introduction The complete severance of a main steamline outside . containment is assumed to occur. The afected SG will rapidly depressurize and release radioiodines initially contained in the secondary coolant and primary coolant activity, transferred via SG tube leaks, directly to the outside atmosphere.* A portion of the iodine activity initially contained in the intact SGs and noble gas activity due to tube leakage is released to atmosphere through eith r the atmospheric dump valves (ADV) or the safety valves (MSSVs). This section describes the assumptions and analyses performed to determine the amount of radioactisity released and the offsite and control room doses resulting from this release.
Impet Parameters and Assumptions -
The analysis of the steam line break (SLB) radiological consequences uses the analytical methods and assunptions outlined in the Standard Review Plan (Reference 1).
l One hundred and two (102) percent of the uprated power level of 1650 MWt (1683 MWt) is used in the analysis. For the pre-accident iodine spike it is assumed that a reactor transient has occurred prior to the
. SLB and has raised the RCS iodine concentration to 60 Ci/gm of dose equivalent (DE) I 131. For the accident initiated iodine spike the reactor trip associated with the SLB creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to the maximum equilibrium RCS Technical Specification concentration of 1.0 pCi/gm of DE l-131. The duration of the accident initiated iodine spike is 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The noble gas activity concentration in the RCS at the time the accident occurs is based on a fuel defect level of 1.0%. This is approximately equal to the Technical Specification value of 100/E bar pCi/gm for :l '
gross radioactivity. The iodine activity concentration of the secondary coolant at the time the SLB occurs is assumed to be equivalent to the Technical Specification limit of 1.2 Ci/gm of DE l 131. The amount of pnmary to secondary SG tube leakage in each of the two SGs is assumed to be equal to the Technical Specification limit for a single SG of 0.35 gpm. No credit for iodine removal is taken for any steam released to the condenser prior to reactor trip and concurrent loss of oKsite power.
The SG connected to the broken steamline is assumed to boil dry within the initial half hour following the i SLB. The entire liquid inventory of this SG is assumed to be steamed off and all of the iodine initially in this SG is released to the emironment. Also, iodine carried over to the faulted SG by SG tube leaks is assumed to be released directly to the emironment with no credit taken for iodine retention in the SG. ;
i An iodine partition factor in the intact SG of 0.01 (curies 1/gm steam)/(curies 1/gm water) is used (Reference 1). All noble gas activity carried over to the secondary through SG tube leakage is assumed to i i . be immediately released to the outside atmosphere. l i
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i Eight hours aAer the accident, the RHR System is assumed to be placed into senice for heat removal, and there are no further steam releases to atmosphere from the secondary system.
The thyroid dose conversion factors, breathing rates, and atmospheric dispersion factors used in the dose calculations are given in Table 1. The core and coolant actisities used in the dose calculations are given in Table 2. The parameters associated with the control room HVAC modes are summarized in Table 3.
The remaining major assumptions and parameters used specifically in this analysis are itemized in Table 10.
CentrolItoom Model The Point Beach control room HVAC system operates in one of four modes Mode 1 is the normal HVAC mode, in which 5% of the air flow is outside air and 95% is recirculated air. Mode 2, which consists of 100% recirculated unfiltered air within the control room, is initiated either by a containment isolation signal or manually from the control room. Mode 3 is initiated manually by operator action and allows for Altered recirculated airflow. Mode 4 is initiated either by a control room radioactisity signal or manually by operator action. In this mode,25% of the available flow is made up with filtered outside air while the-remairdng 75% air flow is unfiltered recirculation. The parameters associated with the control room HVAC modes are summarized in Table 3. These parameters have been taken from Reference 2. In addition, a factor of 10 reduction to the thyroid dose is allowed with the use of Potassium lodide pills by the control room operators.
For the steam line break accident it is assumed that the HVAC system begins in Mode 1. On containment isolation, which is conservatively assumed to begin 5 minutes aAer event initiation, the system is automatically shiAed to Mode 2. When the dose rates in the control room exceed one of the radiation alarm setpoints, the system is automatically shiAed to Mode 4 and remains there until the radiation release has ended. The shiA to Mode 4 will occur within 30 minuter. In order to bound the total dose received by the operators, the radiological consequences were calculated for both Mode 2 and, Mode 4 separately. _ The first case has a shiA from Mode 1 to Mode 2 aAer 5 minutes, and remaining in Mode 2 throughout the event. The second case has a shiA from Mode I to Mode 4 aAct 30 minutes, then remaining in Mode 4 for the rest of the event. The control room doses are calculated over a period of twenty-four hours to ensure that the largest doses to the control room operator are calculated since the ventilation system will continue to operate in the specified modes for several hours following the termination of the steam releases.
Description of Analyses Performed The analysis of the steam line break (SLB) radiological consequences uses the analytical methods and assumptions outlined in the Standard Review Plan (Reference 1). Both the pre-accident iodine spike and acculent initiated iodine spike models are analyzed for these release paths.
Acceptance Criteria The offsite dose limits for a SLB with a pre-accident iodine spike are the guideline values of 10 CFR 100.
These guideline values are 300 rem thyroid and 25 rem y-body. For a SLB with an accident initiated iodine spike the acceptance criteria are a "small fraction of" the 10 CFR 100 guideline values, or 30 rem thyroid and 2.5 rem y body. The criteria defined in SRP Section 6.4 (Reference 3) will be used for the control room dose limits: 30 rem thyroid,5 rem whole body and 30 rem beta skin.
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Results .
The offsite and control room thyroid, y-body, and beta skin doses due to the SLB are given in Table 1 L The results of both control room models have been included.
Conclusions The offsite thyroid and whole bwy oses are within the current NRC ~~:w== criteria for a steamline
- break accident. The control room whole body and beta skin doses are within the current NRC ~~;*==
cntena for the control room. The control room thyroid dose exceeds the 30 rem limit for the cases in which the control room HVAC system continues to operate in Mode 2; however, assuming that the operators would be instructed to take the potassium iodide pills, the control room thyroid dose would be reduced to approximately 14 rem which is within the 30 rem limit. For the cases in which the control room HVAC system switenes to Mode 4 within 30 minutes, the control room thyroid doses are within the NRC e~;*== criteria.
Areas of Conservatism with respect to the Steam Generator Replacement Project
'!his analysis was performed to support both the Steam Generator Replacement Project and the fuel upgrade /uprating program for Point Beach Units 1 and 2. As such, there are several areas in which a more conservative or bounding value was used to support the fuel upgrade or uprating which would not be necessary to support the replacement steam generators alone. This section describes the conservatisms incorporated into the steamline break accident radiological analysis with respect to the Steam Generator !'
Replacement Project.
The most significant impact to the steamline break accident is the use of the analytical methods and assumptions outlined in the Standard Resiew Plan (Reference 1). Specifically, this means the accident was modeled to include both a pre-accident and an accident initiated iodine spike in the coolant actisities where the prior FSAR analysis was based on an equilibrium RCS activity equivalent to 1% fuel defects.
The steam releases for the steamline break accident were calculated using the increased power level of 1650 MWt, a higher Tavg of 580 F, and an RCS pressure of 2250 psi. The steam releases calculat:d with these parameters bound the steam releases which would correspond to a power level of 1520 MWt, a Tavg of 573.9'F and an RCS pressure of 2000 or 2250 psi, in addition, the source term calculations were performed to incorporate the increased core thermal power level and fuel upgrade parameters. The upgraded fuel includes an increase in the mass of the fuel and enrichment which results in an increase to many isotopes in the core and coolant actisities.
References
- 1. NUREG-0800, Standard Review Plan 15.1.5, Appendix, A, " Radiological Cortsequences of Main Steam Line Failures Outside of a PWR, Rev. 2, July 1981.
- 2. Wisconsin Electric letter to NRC, VPNPD-%-099, " Supplement to Technical Specifications Change Requests 188 and 189 Point Beach Nuclear Plant Units 1 and 2," Bob Link, November 20,1996.
- 3. NRC SRP Section 6.4, " Control Room Habitability System", Rev 2, July 1981, NUREG-0800.
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TABLE 1 l
DOSE CONVERSION FACTORS, BREATHING RATES AND ATMOSPHERIC DISPERSION FACTORS l
Isotope Thyroid Dose Conversion Factors "'
(rem /eurie) l 131 1.07 E6 1-132 6.29 E3
'l-133 1.81 E5 1134 1.07 E3 1-135 3.14 E4 Time Period Breathing Rate
(m'/see) 0-8 hr 3.47 E-4 8-24 hr 1.75 E-4 24 720 hr 2.32 E-4 Site Boundary Atmospherie Dispersion Factors
(see/m*)
0-2hr 5.0 E 4 Low Population Zone 0-8 hr 3.0 E-5 8-24 hr 1.6 E 5 24 96 hr 4.2 E-6 96-720 hr 8.6 E-7 Control Room Release from Containment "' Release from Safety Valves "'
0-8 hr 2.1 E-3 1.9 E 3 8 24 hr 1.3 E 3 1.3 E-3 24-96 hr 8.3 E-4 7.6 E-4 96-720 hr 3.3 E-4 2.9 E 4
"'ICRP Publication 30
Regulatory Guide 1.4 0' Wisconsin Electric letter VPNPD-96-099
"' The rod ejection and MSLB release is from containment, the SGTR and locked rotor release is from the safety valves.
. - .-~~-. . -. . . . . - - . - - - . . . - - . .
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TABLE 2 .
CORE AND COOLANT ACTIVITIES
- l Nuclide Total Core Activity at Shutdown Maximum Coolant Actisity (based !
(Cl) on 1% fuel defects)(pCi/gm) ;
l 131 4.4 E7 2.4 EO l 132 6.3 E7 2.4 E0 1-133- 9.0 E7 3.8 E0 1-134 9.9 E7 5.3 E 1 1-135 8.4 E7 1.9 E0 Kr-85 5.4 E5 6.9 E0 Kr-85m 1.2 E7 1.4 EO Kr-87 2.3 E7 9.7 E-I Kr-88 3.2 E7 2.7 EO -
Xe-13 im 4.7 E5 2.5 EO Xe-133 8.9 E7 2.3 E2 Xe-133m 2.8 E6 4.2 EO Xe-135 2.3 E7 7.4 EO Xe-135m 1.7 E7 4.0 E Xe-138 7.5 E7 5.9 E-1
"'These core and coolant activities were specifically recalculated for the Point Beach fuel upgrade /uprating program.
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TABLE 3 CONTROL ROOM PARAMETERS Volume 65,243 ft' Unfiltered Inteakage Mode 1 65.2 cfm Mode 2 65.2 cfm Mode 4 10.0 cfm Normal unfiltered CR llVAC (Mode 1) 1000 cfm Total Flow Rate 19800 cfm Filtered Makeup Mode 2 0cfm Mode 4 4950 cfm Filtered Recirculation 1 l
Mode 2 0cfm '
Mode 4 0 cfm Filter Efficiency Elemental 909'.
Organic 90 %
Particulate 9996 Occupancy Factors 0-1 day 1.0 1 1-4 days 0.6 4-30 days 0.4 i
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TABLE 4 ASSUMPTIONS USED FOR LOCKED ROTOR DOSE ANALYSIS Power (l02%) 1683 MWt Reactor Coolant Noble Gas Activity Prior to Accident 1.0% Fuel Defect Level Reactor Coolant Iodine Activity Prior to Accident 60 pCi/gm of DE l-131 Activity Released to Reactor Coolant from Failed Fuel 100% of Core Gap Activity (Noble Gas & lodine)
Fraction of Core Activity in Gap (Noble Gas & lodine) 0.10 Secondary Coolant Activity Prior to Accident 1.2 pCilgm of DE l-131 Total SG Tube Leak Rate During Accident 0.7 gpm SG lodine Partition Factor 0.01 Duration of Activity Release from Secondary System 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Offsite Power Lost '"
Steam Release from SGs to Environment 206,000 lb (0-2 hr) 434,000 lb (2-8 hr)
J
'" Assumption of a loss of offsite pow er is conservative for the locked rotor dose analysis.
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. - . . - - . . - . - . - - . . . . ~ . . - . . - . = - - .
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TABLES LOCKED ROTOR DOSES i
i Site Boundary (0-2 hr) !
Thyroid 15.6 rem !
y-body 1.8 rem I 4
4 Low Population Zone (0-8 hr) l i Thyroid . 10.0 rem t
't y-body 0.2 rem 11 j Control Room (0-24 hr) l Thyroid 653 rem '"
y-body
(
0.4 rem 4
Beta skin 11.0 rem j 9
4 !
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(" This calculated dose exceeds the 30 rem thyroid limit; however, assuming that the operators would be instructed a
to take the potassium iodide pills, this control room thyroid dose would be reduced to approximately 6.5 rem ,
i which is within the limit.
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TABLE 6 ASSUMPTIONS USED FOR ROD EJECTION ACCIDENT DOSE ANALYSIS Power (102%) 1683 M Wt Reactor Coolant Noble Gas Activity Prior to Accident 1.0% Fuel Defect Level Reactor Coolant lodine Activity Prior to Accident 60 pCilgm of DE l-131 Activity Released to Reactor Coolant AND Containment 10.0% of Core Gap Activity from Failed Fuel (Noble Gas & lodine)
Fraction of Core Activity in Gap (Noble Gas & lodine) 0.10 Activity Released to Reactor Coolant AND Containment J
from Melted Fuel lodine 0.125% of Core Activity Noble Gas 0.25% of Core Activity lodine Removal in Containment instantaneous lodine Plateout 50 %
Secondary Coolant Activity Prior to Accident 1.2 pCilgm of DE l-131 Total SG Tube Leek Rate During Accident 0.35 gpm per SG lodine Partition Factor in SGs 0.01 Containment Free Volume 1.065 x 10' ft' Containment Leak Rate 1 0-24 hr 0.4% / day l
> 24 hr 0.2% / day !
Steam Release from SGs 158,200 lb !
Duration of Steam Release Primary to secondary leakage 1500 seconds initial secondary activity 86 seconds Offsite Power Lost
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TABLE 7 ROD EJECTION OFFSITE & CONTROL ROOM DOSES Site Houndary (0-2 hr)
Thyroid 21.9 rem y-body 0.2 rem Low Population Zone (0-8 hr)
Thyroid 9.4 rem y-body 0.03 rem Control Room (0-24 hr)
Thyroid 122 rem "'
y-body 0.02 rem Beta skin 0.4 rem
"' This calculated dose exceeds the 30 rem thyroid limit; however, assuming that the operators would be instructed to take the potassium iodide pills, this control room thyroid dose would be reduced to approximately 12 rem which is within the limit.
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TABLE 8 i ASSUMPTIONS FOR SGTR DOSE ANALYSIS l
Power (102%) 1683 M Wt f
Reactor Coolant Noble Gas Activity Prior to Accident 1.0% Fuel Defect Level !
Reactor Coolant lodine Activity Prior to Accident Pre-Accident Spike 60 pCi/gm of DE l-131 Accident initiated Spike 1.0 pCi/gm of DE l-131 Reactor Coolant lodine Activity increase Due to 500 times equilibrium release rate from fuel for initial Accident Initiated Spike 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after SGTR I Secondary Coolant Activity Prior to Accident 1.2 pCi/gm of DE l-131 i SG Tube Leak Rate for intact SG During Accident 0.35 gpm Break Flow to Ruptured SG 123,600 lb (0-30 min)
SG lodine Partition Factor 0.01 Duration of Activity Release from Secondary System 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Offsite Power Lost Steam Release from SGs to Environment i
Ruptured SG 74,000 lb (0-30 min)
Intact SC 1,660,000 lb (0-2 hr)'"
1,373,000 lb (2-24 hr) !
'" The actual steam release for 0-2 hours is much lower (232,600 lb); however, this larger value was used in the radiological analysis and is conservative.
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I TABLE 9 SGTR OFFSITE & CONTROL ROOM DOSES 1
Site Boundary (0-2 hr)
Thyroid: Accident Initiated Spike 1.7 rem Thyroid: Pre-Accident Spike 3.5 rem y-body 0.1 rem Low Population Zone (0-8 hr)
Thyroid: Accident initiated Spike 0.1 rem Thyroid: Pre-Accident Spike 0.2 rem j
y-body 0.006 rem i I
Control Room w/ Mode 2 (0-24 hr) i Thyroid: Accident initiated Spike 10.9 rem Thyroid: Pre-Accident Spike 25.2 rem y-body 0.04 rem l
Beta skin 3.6 rem Control Room w/ Mode 4 (0-24 hr)
Thyroid: Accident Initiated Spike 1.4 rem Thyroid: Pre-Accident Spike 3.8 rem y-body 0.005 rem Beta skin 0.3 rem I
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- . e TABLE 10 ASSUMPTIONS USED FOR SLB DOSE ANALYSIS Power (l02%) 1683 M Wt Reactor Coolant Noble Gas Activity Prior to Accident 1.0% Fuel Defect Level Reactor Coolant lodine Activity Prior to Accident Pre-Accident Spike 60 pCi/gm of DE l 131 Accident initiated Spike 1.0 pCi/gm of DE I-131 Reactor Coolant lodine Activity increase Due to 500 times equilibrium release rate from fuel for initial Accident initiated Spike 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after SGTR Secondary Coolant Activity Prior to Accident 1.2 pCilgm of DE l-131 )
SG Tube Leak Rate for intact SG During Accident 0.35 gpm lodine Partition Factor Faulted SG 1.0 (SG assumed to steam dry) l Intact SG 0.01 Duration of Activity Release from Secondary System 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> !
Offsite Power Lost Steam Release from intact SG 212,000 lb (0-2 hr) 405,000 lb (2-8 hr) l l
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TABLE 11 SLB DOSES Site floundary (0-2 hr)
Thyroid: Accident initiated Spike 8.0 rem Thyroid: Pre-Accident Spike 8.3 rem y-body 0.03 rem Low Population Zone (0-8 hr)
Thyroid: Accident Initiated Spike 0.7 rem Thyroid: Pre-Accident Spike 0.7 rem y-body 0.002 rem Control Room w/ Mode 2 (0-24 hr)
Thyroid: Accident Initiated Spike 134 rem '"
Thyroid: Pre-Accident Spike 136 rem '"
y-body 0.006 rem Beta skin 0.08 rem Control Room w/ Mode 4 (0-24 hr)
Thy roid: Accident initiated Spike 15.6 rem Thy roid: Pre-Accident Spike 15.8 rem y-body 0.002 rem Beta skin 0.03 rem J
'" This calculated dose exceeds the 30 rem thyroid limit; however, assuming that the operators would be instructed to take the potassium iodide pills, this control room thyroid dose would be reduced te approximately 14 rem w hich is within the limit.