L-14-245, Response to Request for Additional Information Regarding Request for Licensing Action on Alternative Accident Source Term Radiological Dose Calculations: Difference between revisions
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Page3 As noted above, the dose limit for control room personnel Is specified In 10 CFR Part 50.67 (DIN14). 4.0 METHOD OF ANALYSIS This calculation will evaluate the total effective dose equivalent (TEDE) for the PNPP design basis radiological accident (LOCA) using the revised accident source term based on Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors* (DIN 7). The TEDE dose is defined as the sum of the deep-close equivalent (for external exposures) and the committed effective dose equivalent (for Internal exposures) (DIN 24). The RADTRAD Code, Version 3.03, will be used to calculate radiological consequences in terms of TEDE. RADTRAD <Radionuclide Iransport and Removal .@nd Dose Estimation) calculates fission product transport and removal along with the resulting radiation doses at selected receptors. | Page3 As noted above, the dose limit for control room personnel Is specified In 10 CFR Part 50.67 (DIN14). 4.0 METHOD OF ANALYSIS This calculation will evaluate the total effective dose equivalent (TEDE) for the PNPP design basis radiological accident (LOCA) using the revised accident source term based on Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors* (DIN 7). The TEDE dose is defined as the sum of the deep-close equivalent (for external exposures) and the committed effective dose equivalent (for Internal exposures) (DIN 24). The RADTRAD Code, Version 3.03, will be used to calculate radiological consequences in terms of TEDE. RADTRAD <Radionuclide Iransport and Removal .@nd Dose Estimation) calculates fission product transport and removal along with the resulting radiation doses at selected receptors. | ||
The code is described in NUREG/CR-6604, "A Simplified Model for Radionuclide Transport and Removal and Dose Estimation" (DIN 2, DIN 22, and DIN 23). RADTRAD 3.03 was certified for this application (DIN 30) in accordance with the ENERCON computer code certification procedure | The code is described in NUREG/CR-6604, "A Simplified Model for Radionuclide Transport and Removal and Dose Estimation" (DIN 2, DIN 22, and DIN 23). RADTRAD 3.03 was certified for this application (DIN 30) in accordance with the ENERCON computer code certification procedure | ||
[ENERCON CSP 3.02]. 5.0 ASSUMPTIONS | [ENERCON CSP 3.02]. 5.0 ASSUMPTIONS 5.1 Control Room Emergency Reclrculatlon System (CRERS) Upon receipt of an Engineered Safety Feature (ESF) actuation system signal or high radiation signal, the PNPP control room heating, ventilation, and air conditioning (HVAC) system is designed to automatically isolate and activate the CRERS; however, this analysis conservatively assumes that the normal HVAC system continues to operate with an outside air intake (6000 cfm+10% margin) and exhaust to the environment (4800 cfm) until the CRERS is manually actuated at 30 minutes. Each redundant CRERS subsystem has a high efficiency particulate air (HEPA) filter, charcoal adsorbers and a post HEPA filter. Operation of the CRERS fans, charcoal adsorbers, and HEPA filters are credited in this analysis. | ||
Room Emergency Reclrculatlon System (CRERS) Upon receipt of an Engineered Safety Feature (ESF) actuation system signal or high radiation signal, the PNPP control room heating, ventilation, and air conditioning (HVAC) system is designed to automatically isolate and activate the CRERS; however, this analysis conservatively assumes that the normal HVAC system continues to operate with an outside air intake (6000 cfm+10% margin) and exhaust to the environment (4800 cfm) until the CRERS is manually actuated at 30 minutes. Each redundant CRERS subsystem has a high efficiency particulate air (HEPA) filter, charcoal adsorbers and a post HEPA filter. Operation of the CRERS fans, charcoal adsorbers, and HEPA filters are credited in this analysis. | |||
The CRERS Is an ESF system that Is tested in accordance with R.G. 1.52 (DIN 20). The current test acceptance criterion for the CRERS charcoal adsorbers requires a penetration of less than 2.5% (DIN 21). Based on the testing requirements, a charcoal adsorber removal efficiency of 95% could be justified; however, for additional operational margin, elemental and organic Iodine removal efficiency is assumed to be 80%. Technical Specification 5.5.7 (DIN 21) states that each HEPA filter is tested to show a penetration and system bypass of less than 0.05% when tested in accordance | The CRERS Is an ESF system that Is tested in accordance with R.G. 1.52 (DIN 20). The current test acceptance criterion for the CRERS charcoal adsorbers requires a penetration of less than 2.5% (DIN 21). Based on the testing requirements, a charcoal adsorber removal efficiency of 95% could be justified; however, for additional operational margin, elemental and organic Iodine removal efficiency is assumed to be 80%. Technical Specification 5.5.7 (DIN 21) states that each HEPA filter is tested to show a penetration and system bypass of less than 0.05% when tested in accordance | ||
*with Regulatory Guide 1. 52 (DIN 20). A penetration and bypass of less than 0.05% allows credit for a particulate removal efficiency of 99% per Regulatory Position C.5.c of Regulatory Guide 1.52. This analysis uses a HEPA filter efficiency of 99 percent for aerosol particulates. | *with Regulatory Guide 1. 52 (DIN 20). A penetration and bypass of less than 0.05% allows credit for a particulate removal efficiency of 99% per Regulatory Position C.5.c of Regulatory Guide 1.52. This analysis uses a HEPA filter efficiency of 99 percent for aerosol particulates. | ||
5.2 Hydrogen Mixing System The hydrogen mixing system is manually initiated. | |||
For this analysis, operation of the hydrogen mixing system is not assumed. Due to the minimal (500 cfm) flow rate between the drywall and containment Page4 there is little effect on drywell or containment radionuclide concentrations due to operation of this system. Requiring operators to manually initiate the hydrogen mixing system early in the accident is not a good utilization of operator effort because of the minimal impact on accident doses. 5.3 Control Room lnleakage As described in Section 6.4 of the PNPP USAR (DIN 18), the control room is normally maintained at a slightly positive pressure to the surrounding areas from the 6600 cfm (includes a 10% tolerance on flow rate) fresh air makeup and out leakage of 4800 cfm. In the isolated mode, there is no intake from outside air sources and the control room pressure would eventually reach that of the surrounding areas. After the CRERS is initiated, the maximum control room unfiltered inleakage of 1375 cfm, will be used (DIN 11). The leakage out of the control room envelope is also modeled as 1375 cfm to avoid pressurization of the control room envelope. | |||
Mixing System The hydrogen mixing system is manually initiated. | 5.4 Drywell Flows Leakage from the drywell into the primary containment is due to steaming from the heated reactor core in accordance with R.G. 1.183, Appendix A, Assumption | ||
For this analysis, operation of the hydrogen mixing system is not assumed. Due to the minimal (500 cfm) flow rate between the drywall and containment Page4 there is little effect on drywell or containment radionuclide concentrations due to operation of this system. Requiring operators to manually initiate the hydrogen mixing system early in the accident is not a good utilization of operator effort because of the minimal impact on accident doses. 5.3 Control Room lnleakage As described in Section 6.4 of the PNPP USAR (DIN 18), the control room is normally maintained at a slightly positive pressure to the surrounding areas from the 6600 cfm (includes a 10% tolerance on flow rate) fresh air makeup and out leakage of 4800 cfm. In the isolated mode, there is no intake from outside air sources and the control room pressure would eventually reach that of the surrounding areas. After the CRERS is initiated, the maximum control room unfiltered inleakage of 1375 cfm, will be used (DIN 11). The leakage out of the control room envelope is also modeled as 1375 cfm to avoid pressurization of the control room envelope. | |||
Flows Leakage from the drywell into the primary containment is due to steaming from the heated reactor core in accordance with R.G. 1.183, Appendix A, Assumption | |||
: 3. 7. This leakage is assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early vessel release phases). The termination of the release from the core is due to core recovery and reflood. Instead of evaluating all of the potential steaming rates due to various reflooding scenarios, this analysis will assume that there is a homogenous mixture in the drywell and containment starting at two hours. The assumption of a well-mixed drywall and containment atmosphere at two hours is appropriate because the EAB radiological doses consider the worst two hours as opposed to the first two hours as was done for the previous TIO 14844 source term methodology. | : 3. 7. This leakage is assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early vessel release phases). The termination of the release from the core is due to core recovery and reflood. Instead of evaluating all of the potential steaming rates due to various reflooding scenarios, this analysis will assume that there is a homogenous mixture in the drywell and containment starting at two hours. The assumption of a well-mixed drywall and containment atmosphere at two hours is appropriate because the EAB radiological doses consider the worst two hours as opposed to the first two hours as was done for the previous TIO 14844 source term methodology. | ||
The assumption of a well-mixed drywell and containment atmosphere is implemented by assuming a high mixing flow (2. nE+05 dm, approximately one drywell volume per minute) between the two volumes. The mixing flow is conservatively assumed to continue for the duration of the accident instead of isolating the drywall after the core is quenched. | The assumption of a well-mixed drywell and containment atmosphere is implemented by assuming a high mixing flow (2. nE+05 dm, approximately one drywell volume per minute) between the two volumes. The mixing flow is conservatively assumed to continue for the duration of the accident instead of isolating the drywall after the core is quenched. | ||
The flow rate from the Drywell to the Wetwell, which bypasses the suppression pool, is given in PSAT 150.01C.03 (DIN 11, page 6) as follows: Table 6-1 Drywall and Wetwell Mixing Flows Time After Gap Release Flow from OW to WW Flow from WW to DW (hours) (cfm) (cfm) 0-0.5 0 0 0.5-2.0 3000 0 Pages 2.0-720 2.77E+05 *2.77E+05 | The flow rate from the Drywell to the Wetwell, which bypasses the suppression pool, is given in PSAT 150.01C.03 (DIN 11, page 6) as follows: Table 6-1 Drywall and Wetwell Mixing Flows Time After Gap Release Flow from OW to WW Flow from WW to DW (hours) (cfm) (cfm) 0-0.5 0 0 0.5-2.0 3000 0 Pages 2.0-720 2.77E+05 *2.77E+05 5.5 Containment Leakage Rate The primary containment consists of a drywell, a wetwell and supporting systems to limit fission product leakage during and following the postulated LOCA with rapid isolation of the containment boundary penetrations. | ||
Leakage Rate The primary containment consists of a drywell, a wetwell and supporting systems to limit fission product leakage during and following the postulated LOCA with rapid isolation of the containment boundary penetrations. | |||
The secondary containment will collect and retain fission product leakage from the primary containment and will release fission products to the environment through the Annulus Exhaust Gas Treatment System (AEGTS). During normal operation, the shield building is maintained at a slight negative pressure. | The secondary containment will collect and retain fission product leakage from the primary containment and will release fission products to the environment through the Annulus Exhaust Gas Treatment System (AEGTS). During normal operation, the shield building is maintained at a slight negative pressure. | ||
Following a OBA, it is expected to remain negative, however for a short period it may not be maintained below the design negative pressure value of 0.25-inch. | Following a OBA, it is expected to remain negative, however for a short period it may not be maintained below the design negative pressure value of 0.25-inch. | ||
water gauge (USAR 6.5.3.2.1). | water gauge (USAR 6.5.3.2.1). | ||
Therefore, it will be assumed that the primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA (DINs 11 and 25). However, because the gap release does not begin until two minutes post-accident, this 40-second period when there may be direct release to the environment is not considered. | Therefore, it will be assumed that the primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA (DINs 11 and 25). However, because the gap release does not begin until two minutes post-accident, this 40-second period when there may be direct release to the environment is not considered. | ||
5.6 AEGTS Filtration The AEGTS includes HEPA filters which are periodically tested to demonstrate compliance with Regulatory Guide 1.52. Particulate removal by the HEPA filters is assumed to be 99% in accordance with Regulatory Guide 1.52 (DIN 20). The system also contains 4-inch deep activated charcoal adsorbers to remove elemental and organic iodine; however, this analysis conservatively assumes a removal efficiency of 0% for the charcoal adsorbers to allow operational flexibility. | |||
Filtration The AEGTS includes HEPA filters which are periodically tested to demonstrate compliance with Regulatory Guide 1.52. Particulate removal by the HEPA filters is assumed to be 99% in accordance with Regulatory Guide 1.52 (DIN 20). The system also contains 4-inch deep activated charcoal adsorbers to remove elemental and organic iodine; however, this analysis conservatively assumes a removal efficiency of 0% for the charcoal adsorbers to allow operational flexibility. | |||
The AEGTS extracts and filters a maximum of 2000 cfm from .the annulus. During an accident, the maximum expected discharge to the atmosphere is 1000 cfm (DIN 18). The balance of the filtered AEGTS flow is routed back to the annulus. This analysis conservatively assumes that 2000 cfm Is discharged directly to the environment with no recirculation (holdup) of iodine in the annulus. 5.7 Containment Spray Manual initiation of containment spray is assumed at thirty minutes instead of automatic initiation on high pressures and low water level per previous (DIN 25) analysis. | The AEGTS extracts and filters a maximum of 2000 cfm from .the annulus. During an accident, the maximum expected discharge to the atmosphere is 1000 cfm (DIN 18). The balance of the filtered AEGTS flow is routed back to the annulus. This analysis conservatively assumes that 2000 cfm Is discharged directly to the environment with no recirculation (holdup) of iodine in the annulus. 5.7 Containment Spray Manual initiation of containment spray is assumed at thirty minutes instead of automatic initiation on high pressures and low water level per previous (DIN 25) analysis. | ||
Containment spray is assumed to end at 24 hours at which time the radionuclide removal by containment spray is terminated. | Containment spray is assumed to end at 24 hours at which time the radionuclide removal by containment spray is terminated. | ||
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Therefore, any bypass leakage releases would not be treated by an ESF filtration system prior to being released to the environment. | Therefore, any bypass leakage releases would not be treated by an ESF filtration system prior to being released to the environment. | ||
The secondary containment bypass leakage is currently limited to 5.04% of La. when pressurized to by Technical Specification SR 3.6.1.3.9 (DIN 15) even though the previous LOCA dose calculation (DIN 25) assumed a leakage of 10.08% of L 8* The containment bypass leakage will be maintained at 0.1008 La in this analysis to allow for an increase In the Technical Specifieation allowable leakage limit. 5.11 Source Tenn Release In accordance with R.G. 1.183 (DIN 7), only the gap and in-vessel release phases are considered in this design basis LOCA dose calculation. | The secondary containment bypass leakage is currently limited to 5.04% of La. when pressurized to by Technical Specification SR 3.6.1.3.9 (DIN 15) even though the previous LOCA dose calculation (DIN 25) assumed a leakage of 10.08% of L 8* The containment bypass leakage will be maintained at 0.1008 La in this analysis to allow for an increase In the Technical Specifieation allowable leakage limit. 5.11 Source Tenn Release In accordance with R.G. 1.183 (DIN 7), only the gap and in-vessel release phases are considered in this design basis LOCA dose calculation. | ||
The core source terms are assumed to be released at a constant rate such that the release is completed by the end of the specified release period. Assumptions regarding release fractions and timing are consistent with Tables 1and4 of R.G. 1.183 (DIN 7). Table 1 of R.G. 1.183 Is given below: B'WR Core Inventory Fracdon Released Into Con&alnmeat GroDPl P1ii9Pbase Noble Gases 0.05 0.95 1.0 Halotzens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Telhui1101 Metals 0.00 0.05 0.05 Ba. Sr o.oo 0.02 o.oi Noble Metals 0.00 0.0025 0.0025 Ceri1un Group 0.00 0.0005 0.0005 Lanthanides o.oo 0.0002 0.0002 Table 4 of R.G. 1.183 Is reproduced below: LOCA Release Phases PWRs BWRs Phase Gap Release Early In-Vessel Onset 30 sec O.Shr Duration Onset 0.5 hr 2 min 1.3 hr 0.5 tu* Duration O.Shr 15hr | The core source terms are assumed to be released at a constant rate such that the release is completed by the end of the specified release period. Assumptions regarding release fractions and timing are consistent with Tables 1and4 of R.G. 1.183 (DIN 7). Table 1 of R.G. 1.183 Is given below: B'WR Core Inventory Fracdon Released Into Con&alnmeat GroDPl P1ii9Pbase Noble Gases 0.05 0.95 1.0 Halotzens 0.05 0.25 0.3 Alkali Metals 0.05 0.20 0.25 Telhui1101 Metals 0.00 0.05 0.05 Ba. Sr o.oo 0.02 o.oi Noble Metals 0.00 0.0025 0.0025 Ceri1un Group 0.00 0.0005 0.0005 Lanthanides o.oo 0.0002 0.0002 Table 4 of R.G. 1.183 Is reproduced below: LOCA Release Phases PWRs BWRs Phase Gap Release Early In-Vessel Onset 30 sec O.Shr Duration Onset 0.5 hr 2 min 1.3 hr 0.5 tu* Duration O.Shr 15hr 6.0 Design Input 6.1 Plant Grade The PNPP plant grade elevation is 620 feet (DIN 50). 6.2 Core Source Terms and Releases Pages The PNPP core source term release magnitude, timing and chemical form are based on Regulatory Guide 1.183 (DIN 7). The core source terms were developed by GE Hitachi (DIN 9). The calculated Inventories are based on 2-year GNF2 refueling cycles and serve as input for design basis accident analyses based on Regulatory Guide 1.183 source term assumptions. | ||
Input 6.1 Plant Grade The PNPP plant grade elevation is 620 feet (DIN 50). 6.2 Core Source Terms and Releases Pages The PNPP core source term release magnitude, timing and chemical form are based on Regulatory Guide 1.183 (DIN 7). The core source terms were developed by GE Hitachi (DIN 9). The calculated Inventories are based on 2-year GNF2 refueling cycles and serve as input for design basis accident analyses based on Regulatory Guide 1.183 source term assumptions. | |||
The fission* product inventory calculations were performed using the ORIGEN2 code. The Ci/MN multipliers developed In DIN 9 are applied here to generate the core source terms at the onset of the event. A reactor power level of 3833 MWt will be used based on 102% of the rated thermal power, 3758 MWt, as defined in Technical Specification 1.1, Definitions, page 1.0-5. Amendment 112 (DIN 15). The GNF2 fuel source terms are based on the GNF2 equilibrium source activity given below. The source terms include fission products, actinides, and activation products. | The fission* product inventory calculations were performed using the ORIGEN2 code. The Ci/MN multipliers developed In DIN 9 are applied here to generate the core source terms at the onset of the event. A reactor power level of 3833 MWt will be used based on 102% of the rated thermal power, 3758 MWt, as defined in Technical Specification 1.1, Definitions, page 1.0-5. Amendment 112 (DIN 15). The GNF2 fuel source terms are based on the GNF2 equilibrium source activity given below. The source terms include fission products, actinides, and activation products. | ||
The listing of isotopes given In Table 6-1, below, is based on the isotopes used in the RADTRAD computer code. As stated in the RADTRAD User's Manual, NUREG/CR-6604 (DIN 2), the 60 isotope nuclide file is based on isotopes selected in WASH-1400 | The listing of isotopes given In Table 6-1, below, is based on the isotopes used in the RADTRAD computer code. As stated in the RADTRAD User's Manual, NUREG/CR-6604 (DIN 2), the 60 isotope nuclide file is based on isotopes selected in WASH-1400 | ||
[DIN 53) with the addition of 6 Isotopes used in the MACCS code [DIN 54). Co58 Co60 Kr85 Kr85m Kr87 Kr88 Rb86 Sr89 Sr90 Sr91 Sr92 Y90 Y91 Y92 Y93 Zr95 Table 6-1 Source Term GNF2 Equlllbrium lsotone fCllMWth) 2.647E+02 4.827E+02 3.789E+o2 6.737E+o3 1.283E+o4 1.804E+04 6.882E+01 2.425E+04 3.016E+03 3.064E+04 3.346E+04 3.118E+03 3.152E+04 3.362E+04 3.928E+04 4.440E+04 Isotope Zr97 Nb95 Mo99 Tc99m Ru103 Ru105 Ru106 Rh105 Sb127 Sb129 Te127 Te127m Te129 Te129m Te131m Te132 1131 1132 1133 1134 1135 Xe133 Xe135 Cs134 Cs136 Cs137 Ba139 Ba140 La140 La141 La142 Ce141 Ce143 Ce144 Pr143 Nd147 Np239 Pu238 Pu239 Pu240 GNF2 Equfllbrfum fCllMWth) 4.490E+o4 4.463E+o4 5.105E+04 | [DIN 53) with the addition of 6 Isotopes used in the MACCS code [DIN 54). Co58 Co60 Kr85 Kr85m Kr87 Kr88 Rb86 Sr89 Sr90 Sr91 Sr92 Y90 Y91 Y92 Y93 Zr95 Table 6-1 Source Term GNF2 Equlllbrium lsotone fCllMWth) 2.647E+02 4.827E+02 3.789E+o2 6.737E+o3 1.283E+o4 1.804E+04 6.882E+01 2.425E+04 3.016E+03 3.064E+04 3.346E+04 3.118E+03 3.152E+04 3.362E+04 3.928E+04 4.440E+04 Isotope Zr97 Nb95 Mo99 Tc99m Ru103 Ru105 Ru106 Rh105 Sb127 Sb129 Te127 Te127m Te129 Te129m Te131m Te132 1131 1132 1133 1134 1135 Xe133 Xe135 Cs134 Cs136 Cs137 Ba139 Ba140 La140 La141 La142 Ce141 Ce143 Ce144 Pr143 Nd147 Np239 Pu238 Pu239 Pu240 GNF2 Equfllbrfum fCllMWth) 4.490E+o4 4.463E+o4 5.105E+04 | ||
* 4.470E+o4 4.309E+o4 3.046E+o4 1.750E+04 2.871E+o4 3.016E+03 8.906E+03 2.997E+03 4.049E+02 8.762E+03 1.304E+03 3.965E+03 3.850E+o4 2.714E+o4 3.914E+o4 5.495E+o4 6.025E+o4 5.150E+o4 5.302E+o4 1.934E+o4 6.926E+03 2.162E+03 4.190E+03 4.877E+04 4.709E+o4 5.002E+o4 4.440E+o4 4.278E+o4 4.460E+o4 4.090E+o4 3.670E+o4 3.957E+o4 1.795E+o4 5.619E+05 1.338E+02 1.291E+01 1.749E+01 Pages I 1* I Page10 GNF2 Equlllbrlum tsotoDe (CllMWth) | * 4.470E+o4 4.309E+o4 3.046E+o4 1.750E+04 2.871E+o4 3.016E+03 8.906E+03 2.997E+03 4.049E+02 8.762E+03 1.304E+03 3.965E+03 3.850E+o4 2.714E+o4 3.914E+o4 5.495E+o4 6.025E+o4 5.150E+o4 5.302E+o4 1.934E+o4 6.926E+03 2.162E+03 4.190E+03 4.877E+04 4.709E+o4 5.002E+o4 4.440E+o4 4.278E+o4 4.460E+o4 4.090E+o4 3.670E+o4 3.957E+o4 1.795E+o4 5.619E+05 1.338E+02 1.291E+01 1.749E+01 Pages I 1* I Page10 GNF2 Equlllbrlum tsotoDe (CllMWth) | ||
Pu241 5.748E+03 Am241 7.237E+OO Cm242 1.799E+03 Cm244 1.124E+02 | Pu241 5.748E+03 Am241 7.237E+OO Cm242 1.799E+03 Cm244 1.124E+02 6.2.1 Onset of Gap Release Table 4 of Regulatory Guide 1.183 tabulates values acceptable to the NRC for the onset and duration of each sequential release phase for OBA LOCAs at PWRs and BWRs. The specified onset of the gap release Is the time following the initiation of the accident (i.e., time = 0) prior to the start of the gap release. For a BWR the onset is 2 minutes. Credit will be taken for decay prior to the onset of the gap release at 2 minutes. 6.2.2 Release fractions The release fractions used in this analysis are consistent with Table 1 of R.G. 1.183 (DIN 7) which is reproduced in Section 5.11 of this calculation. | ||
6.3 Suppression Pool Iodine Re-evolution The impact of any postulated iodine re-evolution from the suppression pool has been evaluated and shown to be negligible based on the pool pH level. If the pH is maintained above 7, very little (less than 1 %) of the dissolved Iodine will be converted to elemental iodine (DIN 1, DIN 7). The Standby Liquid Control System (SLCS) is used for controlling and maintaining long-term suppression pool water pH levels to 7 or above. The pH of post-accident water in the containment will remain above 7 for the entire duration of the postulated LOCA (DIN 26). As such, this analysis will not consider any impad to the offsite or control room doses due to iodine re-evolution from the suppression pool. Also, in accordance With Appendix A of Regulatory Guide 1.183 (DIN 7), because the suppression pool pH is controlled at values of 7 or greater, the chemical form of radloiodlne released to the containment can be assumed to be 95% cesium Iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. 6.4 Dose Conversion Factors The effective dose conversion factors for the TEDE calculations are based on FGR 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (DIN 4) and FGR 12, "External Exposure to Radlonuclldes In Air, Water, and Soiin (DIN 5). These reports tabulate dose coefficients for external exposure to photons and electrons emitted by radionuclides distributed in air, water, and soil, as well as, dose coefficients for the committed dose equivalent to tissues of the body per unit activity of inhaled or ingested radlonuclides. | |||
of Gap Release Table 4 of Regulatory Guide 1.183 tabulates values acceptable to the NRC for the onset and duration of each sequential release phase for OBA LOCAs at PWRs and BWRs. The specified onset of the gap release Is the time following the initiation of the accident (i.e., time = 0) prior to the start of the gap release. For a BWR the onset is 2 minutes. Credit will be taken for decay prior to the onset of the gap release at 2 minutes. 6.2.2 Release fractions The release fractions used in this analysis are consistent with Table 1 of R.G. 1.183 (DIN 7) which is reproduced in Section 5.11 of this calculation. | |||
Pool Iodine Re-evolution The impact of any postulated iodine re-evolution from the suppression pool has been evaluated and shown to be negligible based on the pool pH level. If the pH is maintained above 7, very little (less than 1 %) of the dissolved Iodine will be converted to elemental iodine (DIN 1, DIN 7). The Standby Liquid Control System (SLCS) is used for controlling and maintaining long-term suppression pool water pH levels to 7 or above. The pH of post-accident water in the containment will remain above 7 for the entire duration of the postulated LOCA (DIN 26). As such, this analysis will not consider any impad to the offsite or control room doses due to iodine re-evolution from the suppression pool. Also, in accordance With Appendix A of Regulatory Guide 1.183 (DIN 7), because the suppression pool pH is controlled at values of 7 or greater, the chemical form of radloiodlne released to the containment can be assumed to be 95% cesium Iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. 6.4 Dose Conversion Factors The effective dose conversion factors for the TEDE calculations are based on FGR 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (DIN 4) and FGR 12, "External Exposure to Radlonuclldes In Air, Water, and Soiin (DIN 5). These reports tabulate dose coefficients for external exposure to photons and electrons emitted by radionuclides distributed in air, water, and soil, as well as, dose coefficients for the committed dose equivalent to tissues of the body per unit activity of inhaled or ingested radlonuclides. | |||
These dose coefficients for exposure to radiation are intended for the use in calculating the dose equivalent to organs and tissues of the body and are endorsed by the NRC in Regulatory Guide 1.183, Sections 4.1.2 (FGR Page 11 11) and 4.1.4 (FGR 12). Dose conversion factors for the 60-isotope, 9 element NUREG 1465 {DIN 3) accident source term composition are included in the RADTRAD Input. 6.5 Atmospheric Dispersion Factors The atmospheric dispersion factors ('xjQ values) for the LPZ and EAB are obtained from PSAT 04202U.03 (DIN 13, page 10) and Calculation 3.2.6.3 (DIN 46). The Control Room atmospheric dispersion factors are documented in DES/98-845 (DIN 10). The Technical Support Center atmospheric dispersion factors are documented in PY-NUS/CEl-1474 (DIN 48) and Revision 0 of the TSC Dose Evaluation (DIN 49). The x/Q values, based on a ground level release, are given below: Table&-3 'IJQ (seclm 3) Location* | These dose coefficients for exposure to radiation are intended for the use in calculating the dose equivalent to organs and tissues of the body and are endorsed by the NRC in Regulatory Guide 1.183, Sections 4.1.2 (FGR Page 11 11) and 4.1.4 (FGR 12). Dose conversion factors for the 60-isotope, 9 element NUREG 1465 {DIN 3) accident source term composition are included in the RADTRAD Input. 6.5 Atmospheric Dispersion Factors The atmospheric dispersion factors ('xjQ values) for the LPZ and EAB are obtained from PSAT 04202U.03 (DIN 13, page 10) and Calculation 3.2.6.3 (DIN 46). The Control Room atmospheric dispersion factors are documented in DES/98-845 (DIN 10). The Technical Support Center atmospheric dispersion factors are documented in PY-NUS/CEl-1474 (DIN 48) and Revision 0 of the TSC Dose Evaluation (DIN 49). The x/Q values, based on a ground level release, are given below: Table&-3 'IJQ (seclm 3) Location* | ||
Time Interval EAB LPZ Oto 2 hrs 4.3E-4 4.8E-5 2to8hrs 4.8E-5 8to 24 hrs 3.3E-5 24to 96 hrs 1.4E-5 96to 720 hrs 4.1E-6 Table 6-4 Control Room and TSC 'IJQ (sec/m 3) Time Interval CONTROL ROOM TSC Oto 8 hrs 3.5E-4 5.1E-5 8 to 24 hrs 2.1E-4 4.1E-5 24 to 96 hrs 1.1E-4 3.1E-5 96to720 hrs 5.75E-5 1.1E-5 6.6 Breathing Rate and Occupancy Factors The breathing rates applied in the calculation of the Inhalation dose are consistent with those reported In Sections 4.1.3 and 4.2.6 of R.G. 1.183 (DIN 7). Time Period Oto 8 hours Bto 24 hours 1to30 days . ' Table&-5 Breathing Rates (m 3/s) EAB LPZ 3.5E-4 3.5E-4 1.BE-4 1.8E-4 2.3E-4 2.3E-4 Control Room/TSC 3.5E-4 3.5E-4 3.5E-4 Page 12 The. control room and TSC occupancy factor$ are consistent with those reported in Section 4.2.6 of R.G. 1.183 and are tabulated below. | Time Interval EAB LPZ Oto 2 hrs 4.3E-4 4.8E-5 2to8hrs 4.8E-5 8to 24 hrs 3.3E-5 24to 96 hrs 1.4E-5 96to 720 hrs 4.1E-6 Table 6-4 Control Room and TSC 'IJQ (sec/m 3) Time Interval CONTROL ROOM TSC Oto 8 hrs 3.5E-4 5.1E-5 8 to 24 hrs 2.1E-4 4.1E-5 24 to 96 hrs 1.1E-4 3.1E-5 96to720 hrs 5.75E-5 1.1E-5 6.6 Breathing Rate and Occupancy Factors The breathing rates applied in the calculation of the Inhalation dose are consistent with those reported In Sections 4.1.3 and 4.2.6 of R.G. 1.183 (DIN 7). Time Period Oto 8 hours Bto 24 hours 1to30 days . ' Table&-5 Breathing Rates (m 3/s) EAB LPZ 3.5E-4 3.5E-4 1.BE-4 1.8E-4 2.3E-4 2.3E-4 Control Room/TSC 3.5E-4 3.5E-4 3.5E-4 Page 12 The. control room and TSC occupancy factor$ are consistent with those reported in Section 4.2.6 of R.G. 1.183 and are tabulated below. | ||
* Table 6-6 Control Room and TSC Occupancy Factors Time Period Occupancy Factor Oto 24 hours 1.0 1 to4 days 0.6 4to30 days 0.4 6. 7 Containment Volumes The volumes of the containment regions are from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31). Table 6*7 Containment Volumes Region Volume (ft"') Unsprayed Containment 684,226 Sprayed Containment 481,174 Drywall Note: the above volumes are shown as rounded values in the RADTRAD screen views but the actual values are used in the RADTRAD input file. 6.8 Technical Support Center Doses Doses to personnel in the Technical Support Center (TSC) are calculated in the same manner as the doses to the Control Room operators except for the TSC specific atmospheric dispersion, xfQ, values and TSC data. The additional data needed to determine the TSC doses is as follows (DIN 49, page 9 and 10, and DI 5.7.1 page 18): Parameter TSC Volume (ff) HVAC Flow Rate (cfm) Table&-8 TSCData Recirculation Filter Flow Rate (cfm) Charcoal Filter Bed Depth (in) Filtered Damper lnleakage (cfm)* Unfiltered lnleakage (cfm)** *Added to unfiltered inleakage Value 113,412 37,000 6,000 2 12 27.2 Page 13 -After recirculation is initiated at 60 minutes (includes 1 O cfm for ingress and egress) For the TSC charcoal removal efficiency, a removal efficiency of 80% will be used to provide margin as was done for the Control Room charcoal adsorber removal efficiency. | * Table 6-6 Control Room and TSC Occupancy Factors Time Period Occupancy Factor Oto 24 hours 1.0 1 to4 days 0.6 4to30 days 0.4 6. 7 Containment Volumes The volumes of the containment regions are from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31). Table 6*7 Containment Volumes Region Volume (ft"') Unsprayed Containment 684,226 Sprayed Containment 481,174 Drywall Note: the above volumes are shown as rounded values in the RADTRAD screen views but the actual values are used in the RADTRAD input file. 6.8 Technical Support Center Doses Doses to personnel in the Technical Support Center (TSC) are calculated in the same manner as the doses to the Control Room operators except for the TSC specific atmospheric dispersion, xfQ, values and TSC data. The additional data needed to determine the TSC doses is as follows (DIN 49, page 9 and 10, and DI 5.7.1 page 18): Parameter TSC Volume (ff) HVAC Flow Rate (cfm) Table&-8 TSCData Recirculation Filter Flow Rate (cfm) Charcoal Filter Bed Depth (in) Filtered Damper lnleakage (cfm)* Unfiltered lnleakage (cfm)** *Added to unfiltered inleakage Value 113,412 37,000 6,000 2 12 27.2 Page 13 -After recirculation is initiated at 60 minutes (includes 1 O cfm for ingress and egress) For the TSC charcoal removal efficiency, a removal efficiency of 80% will be used to provide margin as was done for the Control Room charcoal adsorber removal efficiency. | ||
Note that in the RADTRAD files, the TSC is labeled as the Control Room. The normal HVAC flow is assumed to operate for the first 60 minutes after which It is isolated and the recirculation filter is initiated. | Note that in the RADTRAD files, the TSC is labeled as the Control Room. The normal HVAC flow is assumed to operate for the first 60 minutes after which It is isolated and the recirculation filter is initiated. | ||
6.9 Mixing Between the Unsprayed Containment and Sprayed The mixing rate between the unsprayed containment and the sprayed containment was determined to be 71,400 dm in Calculation PSAT 04202U.03, Rev. 0 (DIN 13, page 6). 6.1 O Containment Leakrate The maximum allowable primary containment leakrate, La, is 0.2 volume percent per day at the peak containment pressure (Pa) of7.80 psig per Technical Specification 5.5.12 (DIN 15). Per Assumption 5.5, primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA, when the shield building may not be at a negativ_e pressure. | |||
Between the Unsprayed Containment and Sprayed The mixing rate between the unsprayed containment and the sprayed containment was determined to be 71,400 dm in Calculation PSAT 04202U.03, Rev. 0 (DIN 13, page 6). 6.1 O Containment Leakrate The maximum allowable primary containment leakrate, La, is 0.2 volume percent per day at the peak containment pressure (Pa) of7.80 psig per Technical Specification 5.5.12 (DIN 15). Per Assumption 5.5, primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA, when the shield building may not be at a negativ_e pressure. | |||
This potential 40 second release directly to the environment is not modeled because it has no dose significance. | This potential 40 second release directly to the environment is not modeled because it has no dose significance. | ||
Following this 40-second period, the annulus will collect and retain any fission produd leakage from the primary containment and will release fission products to the environment through the AEGTS. Secondary Containment Bypass leakage is a portion of the total containment leakage, La. Technical Specification SR 3.6.1.3.9 (DIN 15) limits the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. The containment bypass leakage for this calculation Is assumed to be 0.1008 La. (Assumption 5.10) Therefore, the leakrate for the sprayed and unsprayed (including drywell) regions of the containment is calculated below: * [(4.812e + 05 ft 3) | Following this 40-second period, the annulus will collect and retain any fission produd leakage from the primary containment and will release fission products to the environment through the AEGTS. Secondary Containment Bypass leakage is a portion of the total containment leakage, La. Technical Specification SR 3.6.1.3.9 (DIN 15) limits the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. The containment bypass leakage for this calculation Is assumed to be 0.1008 La. (Assumption 5.10) Therefore, the leakrate for the sprayed and unsprayed (including drywell) regions of the containment is calculated below: * [(4.812e + 05 ft 3) | ||
Line 403: | Line 379: | ||
curve contained in the EOP. The containment pressure threshold is met within 30 seconds of the LOCA per DIN 27. 122 minutes to 24 hours The source term release from the vessel is terminated at 2 hours after the onset of gap release with the actuation of ECCS, which results in large amounts of steam evolution and large flows out of the drywall into the containment. | curve contained in the EOP. The containment pressure threshold is met within 30 seconds of the LOCA per DIN 27. 122 minutes to 24 hours The source term release from the vessel is terminated at 2 hours after the onset of gap release with the actuation of ECCS, which results in large amounts of steam evolution and large flows out of the drywall into the containment. | ||
The drywall and lower containment region are assumed to become well-mixed at 2 hours./ ,A,. ,.1.1*1'f | The drywall and lower containment region are assumed to become well-mixed at 2 hours./ ,A,. ,.1.1*1'f | ||
* 24 hours to 30 days Releases to the environment via the containment bypass, MSIV leakage, ECCS leakage and AEGTS exhaust continue for 30 days. As discussed above, containment leakage is reduced at 24 hours based on containment pressure. | * 24 hours to 30 days Releases to the environment via the containment bypass, MSIV leakage, ECCS leakage and AEGTS exhaust continue for 30 days. As discussed above, containment leakage is reduced at 24 hours based on containment pressure. | ||
8.0 MODEL DEVELOPMENT This analysis considers the following three pathways through which source terms can be released from the containment | |||
DEVELOPMENT This analysis considers the following three pathways through which source terms can be released from the containment | |||
* . ECCS liquid leakage outside of containment | * . ECCS liquid leakage outside of containment | ||
* MSIV leakage | * MSIV leakage | ||
Line 422: | Line 396: | ||
The volume of the two intact steam lines between the reactor vessel and the inboard MSIVs is 440 ft3 (DIN 25). The leakage past the first MSIVs in these lines is released to the volume between the first and second MSIVs which Is 292 ft3, two times the volume between the MSIVs In one steam line (146 ft3) (DIN 11). Leakage past the second MSIVs In these lines is also released directly to | The volume of the two intact steam lines between the reactor vessel and the inboard MSIVs is 440 ft3 (DIN 25). The leakage past the first MSIVs in these lines is released to the volume between the first and second MSIVs which Is 292 ft3, two times the volume between the MSIVs In one steam line (146 ft3) (DIN 11). Leakage past the second MSIVs In these lines is also released directly to | ||
* the environment. | * the environment. | ||
This configuration was previously identified in DIN 17 to be limiting with respect to dose consequences. | This configuration was previously identified in DIN 17 to be limiting with respect to dose consequences. | ||
8.2.3 Flows This calculation will apply a maximum MSIV leak rate of 250 scfh with the worst-case main steam line leaking no more than 100 scfh. The leakage limit is assumed to occur: (1) 100 scfh through the broken steam line, (2) 100 scfh through a second intact steam line, and (3) the remaining 50 scfh through a third intact steam line. As stated above, leakage is modeled to occur through two paths, one path consisting of the broken steam line and a second path consisting of the second and third Intact steam lines. All leakage past the outboard MSIVs is assumed to be released to the environment. | |||
This calculation will apply a maximum MSIV leak rate of 250 scfh with the worst-case main steam line leaking no more than 100 scfh. The leakage limit is assumed to occur: (1) 100 scfh through the broken steam line, (2) 100 scfh through a second intact steam line, and (3) the remaining 50 scfh through a third intact steam line. As stated above, leakage is modeled to occur through two paths, one path consisting of the broken steam line and a second path consisting of the second and third Intact steam lines. All leakage past the outboard MSIVs is assumed to be released to the environment. | |||
The drywell atmosphere will not be at standard conditions after the reactor blowdown. | The drywell atmosphere will not be at standard conditions after the reactor blowdown. | ||
The MSIV leakage rate must be.converted to a flow at the drywell conditions. | The MSIV leakage rate must be.converted to a flow at the drywell conditions. | ||
Line 436: | Line 408: | ||
& Containment Bypass Leakage 8.3.1 | & Containment Bypass Leakage 8.3.1 | ||
* Volumes In addition to the main steam lines, the following volumes are used in the LOCA airborne leakage dose calculation (DIN 11 and DIN 31): Volume Name In In Model Model 1 Drvwell 5 Sprayed 6 Unsprayed 7 Annulus 9 Control Room Orywell Table 8-2 LOCA Volumes Description Sprayed Region of the Containment above the Operating Floor at El. 208'10n Unsprayed Reaion of the Containment Secondary Containment Control Room Volume (tr') 2.765E+05 4.812E+05 6.842E+05 1.96E+05 390,020 The volume of the Perry control room has recently been re-evaluated. | * Volumes In addition to the main steam lines, the following volumes are used in the LOCA airborne leakage dose calculation (DIN 11 and DIN 31): Volume Name In In Model Model 1 Drvwell 5 Sprayed 6 Unsprayed 7 Annulus 9 Control Room Orywell Table 8-2 LOCA Volumes Description Sprayed Region of the Containment above the Operating Floor at El. 208'10n Unsprayed Reaion of the Containment Secondary Containment Control Room Volume (tr') 2.765E+05 4.812E+05 6.842E+05 1.96E+05 390,020 The volume of the Perry control room has recently been re-evaluated. | ||
The current volume to be used in Control Room Dose calculations is 390,020 ft3 (DIN 16). | The current volume to be used in Control Room Dose calculations is 390,020 ft3 (DIN 16). | ||
8.3.2 Flows From Drywell Volume Into Containment (Suppression Pool Bypass) The flow rate from the Drywell to the Wetwell are given below, see PSAT 04212H.02 (DIN 43): Time After Gap Release (hours) 0-0.5 0.5-2.0 2.0-720 TableS-3 Drywall Flows Flow from OW to WW (cfm) 0 3000 2.77E+05 Flow from WW to OW (cfm) 0 0 2.77E+05 Page 30 At two hours, the drywell and unsprayed portion of the containment will be assumed to become instantly well-mixed without credit for suppression pool scrubbing in accordance with Regulatory Guide 1.183, Section 3.7 (DIN 7). From Unsprayed and Sprayed Containment Volumes to the Environment By the time the gap release* begins, the containment is completely isolated and only containment leakage is assumed. The design basis containment leakage for Perry is 0.2% per day. Since the AEGTS will not completely draw down the annulus for 40 seconds, a 40 second pressure period is assumed in which all containment leakage is assumed to leak directly to the environment, but because there is no radionuclide release during this 40 second time period, before gap release which begins at two minutes, this leakage does not contribute to the onsite or offsite doses and is not included in the model. From Unsprayed and Sprayed Containment Volumes to the Annulus The majority (89.92%) of the total containment leakage (la) is drawn into the annulus by the AEGTS. Although the primary containment is enclosed by the secondary containment, there are systems that penetrate both the primary containment and the shield building boundaries that could create potential pathways through which fission products in the primary containment could bypass the leakage collection and filtration systems associated with the shield building. | |||
From Drywell Volume Into Containment (Suppression Pool Bypass) The flow rate from the Drywell to the Wetwell are given below, see PSAT 04212H.02 (DIN 43): Time After Gap Release (hours) 0-0.5 0.5-2.0 2.0-720 TableS-3 Drywall Flows Flow from OW to WW (cfm) 0 3000 2.77E+05 Flow from WW to OW (cfm) 0 0 2.77E+05 Page 30 At two hours, the drywell and unsprayed portion of the containment will be assumed to become instantly well-mixed without credit for suppression pool scrubbing in accordance with Regulatory Guide 1.183, Section 3.7 (DIN 7). From Unsprayed and Sprayed Containment Volumes to the Environment By the time the gap release* begins, the containment is completely isolated and only containment leakage is assumed. The design basis containment leakage for Perry is 0.2% per day. Since the AEGTS will not completely draw down the annulus for 40 seconds, a 40 second pressure period is assumed in which all containment leakage is assumed to leak directly to the environment, but because there is no radionuclide release during this 40 second time period, before gap release which begins at two minutes, this leakage does not contribute to the onsite or offsite doses and is not included in the model. From Unsprayed and Sprayed Containment Volumes to the Annulus The majority (89.92%) of the total containment leakage (la) is drawn into the annulus by the AEGTS. Although the primary containment is enclosed by the secondary containment, there are systems that penetrate both the primary containment and the shield building boundaries that could create potential pathways through which fission products in the primary containment could bypass the leakage collection and filtration systems associated with the shield building. | |||
The Perry Technical Specification SR 3.6.1.3.9 (DIN 15) limit the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. This analysis uses a bypass leakrate of 10.08 percent of the primary containment leak rate. . From Unsprayed and Sprayed Containment Volumes to the Annulus As stated above, the majority (89.92%) of the total containment leakage (la) is drawn into the annulus where it is filtered by the installed HEPA filters at a credited efficiency of 99% before being released Into the environment. | The Perry Technical Specification SR 3.6.1.3.9 (DIN 15) limit the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. This analysis uses a bypass leakrate of 10.08 percent of the primary containment leak rate. . From Unsprayed and Sprayed Containment Volumes to the Annulus As stated above, the majority (89.92%) of the total containment leakage (la) is drawn into the annulus where it is filtered by the installed HEPA filters at a credited efficiency of 99% before being released Into the environment. | ||
Page31 Mixing Between the Unsprayed Containment and Sprayed Containment The mixing rate between the unsprayed containment and the sprayed containment is 71,400 cfm, Calculation PSAT 04202U.03, Rev. 0(DIN13). | Page31 Mixing Between the Unsprayed Containment and Sprayed Containment The mixing rate between the unsprayed containment and the sprayed containment is 71,400 cfm, Calculation PSAT 04202U.03, Rev. 0(DIN13). | ||
From Secondary Containment The only flow from secondary containment is via the AEGTS system which draws 2000 cfm through a charcoal-filter unit and HEPA filter unit. The HEPA filters are tested per Regulatory Guide 1.52 and therefore are credited for a 99% removal efficiency in the analysis; however, no credit is taken for the charcoal adsorbers in this analysis. | From Secondary Containment The only flow from secondary containment is via the AEGTS system which draws 2000 cfm through a charcoal-filter unit and HEPA filter unit. The HEPA filters are tested per Regulatory Guide 1.52 and therefore are credited for a 99% removal efficiency in the analysis; however, no credit is taken for the charcoal adsorbers in this analysis. | ||
8.3.3 Removal Mechanisms Natural removal mechanisms for elemental iodine and aerosols will be applied in this calculation using NRC correlations. | |||
Mechanisms Natural removal mechanisms for elemental iodine and aerosols will be applied in this calculation using NRC correlations. | |||
Elemental iodine removal is credited in the drywell and containment volumes. Aerosol removal is credited only in the drywall and unsprayed region of the containment since containment spray will adversely impact the particle size distribution In the containment. | Elemental iodine removal is credited in the drywell and containment volumes. Aerosol removal is credited only in the drywall and unsprayed region of the containment since containment spray will adversely impact the particle size distribution In the containment. | ||
Fission product removal by containment sprays is considered. | Fission product removal by containment sprays is considered. | ||
Line 452: | Line 420: | ||
Decontamination Factor Reductions As discussed above, the elemental iodine removal by natural deposition is neglected after a OF of.200 is reached. Based on the elemental Iodine lambda of 0.878 hf 1 in the drywell, a OF of 200 would be reached in approximately six hours without any leakage or decay. The output in Attachment 2 indicates that the Orywell 2-hour post-accident release (I.e., 2.0000 hr) elemental 1-131 inventory of 2.4087E+22 atoms has reduced to 1.29E+20 at 3.0 hours post-accident representing a OF of 186. This calculation will therefore model the elemental Iodine removal to end at 3.0 hours in the drywell. In the containment, the total (sprayed+ | Decontamination Factor Reductions As discussed above, the elemental iodine removal by natural deposition is neglected after a OF of.200 is reached. Based on the elemental Iodine lambda of 0.878 hf 1 in the drywell, a OF of 200 would be reached in approximately six hours without any leakage or decay. The output in Attachment 2 indicates that the Orywell 2-hour post-accident release (I.e., 2.0000 hr) elemental 1-131 inventory of 2.4087E+22 atoms has reduced to 1.29E+20 at 3.0 hours post-accident representing a OF of 186. This calculation will therefore model the elemental Iodine removal to end at 3.0 hours in the drywell. In the containment, the total (sprayed+ | ||
unsprayed) elemental 1-131 inventory Is 3.1526E+21 atoms after the drywell is flushed at two hours post-accident (i.e., 2.0000 hr). The total activity in both regions of the Page 32 containment is considered because, if only the activity in the sprayed region of the containment Yias considered, a longer period of the higher lambdas would be applicable. | unsprayed) elemental 1-131 inventory Is 3.1526E+21 atoms after the drywell is flushed at two hours post-accident (i.e., 2.0000 hr). The total activity in both regions of the Page 32 containment is considered because, if only the activity in the sprayed region of the containment Yias considered, a longer period of the higher lambdas would be applicable. | ||
At 3. 75 hours, the total elemental 1-131 Inventory in the sprayed and unsprayed containment regions is 1.6644E+19 atoms, representing a OF of 189. This calculation will model spray removal to end at 3. 75 hours. The particulate removal (Powers Model) in the sprayed region of the containment is reduced by a factor of 10 when the aerosol activity is reduced by a OF of 50. In the containment, the total (sprayed + unsprayed) particulate inventory is 8.13 kg after the drywall is flushed at two hours. At 4.95 hours, the total aerosol inventory Is 1.64E-01 kg, representing a OF of 49.7. This calculation will model this removal coefficient to be reduced at 4.95 hours. This is accomplished t>Y reducing the spray flow used in the Powers Model by a factor of ten at this time. 8.3.4 Release Points All source terms released via containment leakage are released through the plant vent 8.3.5 Model The containment airborne model is illustrated In Figure 2 which is based on the time at which the gap release begins. This figure also includes the MSIV leakage transport pathways. | At 3. 75 hours, the total elemental 1-131 Inventory in the sprayed and unsprayed containment regions is 1.6644E+19 atoms, representing a OF of 189. This calculation will model spray removal to end at 3. 75 hours. The particulate removal (Powers Model) in the sprayed region of the containment is reduced by a factor of 10 when the aerosol activity is reduced by a OF of 50. In the containment, the total (sprayed + unsprayed) particulate inventory is 8.13 kg after the drywall is flushed at two hours. At 4.95 hours, the total aerosol inventory Is 1.64E-01 kg, representing a OF of 49.7. This calculation will model this removal coefficient to be reduced at 4.95 hours. This is accomplished t>Y reducing the spray flow used in the Powers Model by a factor of ten at this time. 8.3.4 Release Points All source terms released via containment leakage are released through the plant vent 8.3.5 Model The containment airborne model is illustrated In Figure 2 which is based on the time at which the gap release begins. This figure also includes the MSIV leakage transport pathways. | ||
8.4 Control Room Although the current configuration of the control room HVAC system would automatically Initiate the control room recirculation on a LOCA signal, this analysis assumes that the CRERS is manually initiated at thirty minutes. Once the CRERS is initiated, CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA filters and charcoal adsorbers before being returned to the control room. The normal control room recirculation air flow is 45,000 cfm (DIN 52) including 6,600 cfm of outside air for ventilation. | |||
Room Although the current configuration of the control room HVAC system would automatically Initiate the control room recirculation on a LOCA signal, this analysis assumes that the CRERS is manually initiated at thirty minutes. Once the CRERS is initiated, CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA filters and charcoal adsorbers before being returned to the control room. The normal control room recirculation air flow is 45,000 cfm (DIN 52) including 6,600 cfm of outside air for ventilation. | |||
To represent the normal positive pressurization in the control room, the exfiltration air flow is modeled as 4,600 cfm before Isolation at 30 minutes. The RADTRAD code only allows a single control room recirculation air flow. As a result, the normal recirculation air flow Is not modeled. This is acceptable because the normal recirculation flow does not change the radionuclide concentration in the control room. After Isolation, unfiltered inleakage of 1375 cfm is assumed to be drawn from the control room intake duct for the duration of the postulated accident (30 days). The flow from the control room to the environment is also set at 1375 cfm to avoid pressurization. | To represent the normal positive pressurization in the control room, the exfiltration air flow is modeled as 4,600 cfm before Isolation at 30 minutes. The RADTRAD code only allows a single control room recirculation air flow. As a result, the normal recirculation air flow Is not modeled. This is acceptable because the normal recirculation flow does not change the radionuclide concentration in the control room. After Isolation, unfiltered inleakage of 1375 cfm is assumed to be drawn from the control room intake duct for the duration of the postulated accident (30 days). The flow from the control room to the environment is also set at 1375 cfm to avoid pressurization. | ||
Consistent with the requirements of R. G. 1.183, the contribution to the control room dose due to shine from the containment building (0.13 Rem) and release plume (0.002 Rem) must be considered. | Consistent with the requirements of R. G. 1.183, the contribution to the control room dose due to shine from the containment building (0.13 Rem) and release plume (0.002 Rem) must be considered. | ||
Line 461: | Line 427: | ||
These Page 33 assumptions are conseNative compared to the Alternative Source Term methodology (AST) due to removal of radlolodines from the containment atmosphere by sprays and deposition thereby reducing the radionuclide concentration in containment. | These Page 33 assumptions are conseNative compared to the Alternative Source Term methodology (AST) due to removal of radlolodines from the containment atmosphere by sprays and deposition thereby reducing the radionuclide concentration in containment. | ||
In addition, the total halogen release fraction is 0.3 for the | In addition, the total halogen release fraction is 0.3 for the | ||
* AST methodology providing additional margin. Based on these considerations, the previously calculated shine and plume doses are considered bounding for this analysis. | * AST methodology providing additional margin. Based on these considerations, the previously calculated shine and plume doses are considered bounding for this analysis. | ||
8.5 RADTRAD MODEL The models developed for the analysis are illustrated in Figures 1 and 2. 9.0 Operator Actions The operator actions assumed in this analysis include the following: | |||
MODEL The models developed for the analysis are illustrated in Figures 1 and 2. 9.0 Operator Actions The operator actions assumed in this analysis include the following: | |||
: 1. Manual initiation of containment spray at 30 minutes 2. Manual initiation of CRERS at 30 minutes 3. The pH calculation (DIN 26) assumes initiation of Standby Liquid Control (681:6) to control suppression pool pH SL r.. I&.> I. i1*1'1 SUPPRESSION POOL-1 114,379 ft 3 Figure 1 Fission Product Transport Model (ECCS Leakage Pathway) 1-Pool to Environment Page 34 0.03 2 s 24.5-end} | : 1. Manual initiation of containment spray at 30 minutes 2. Manual initiation of CRERS at 30 minutes 3. The pH calculation (DIN 26) assumes initiation of Standby Liquid Control (681:6) to control suppression pool pH SL r.. I&.> I. i1*1'1 SUPPRESSION POOL-1 114,379 ft 3 Figure 1 Fission Product Transport Model (ECCS Leakage Pathway) 1-Pool to Environment Page 34 0.03 2 s 24.5-end} | ||
ENVIRONMENT -2 6. 71 cfm (24-24.5 hrs) 2 -Control Room Intake . 6000 cfm + 10% = 6600 cfm 1375 cfm after CRERS initiation CONTROL ROOM -3 390,020tt3 | ENVIRONMENT -2 6. 71 cfm (24-24.5 hrs) 2 -Control Room Intake . 6000 cfm + 10% = 6600 cfm 1375 cfm after CRERS initiation CONTROL ROOM -3 390,020tt3 | ||
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None DESCRIBE WHERE THE ADDENDUM WILL BE EVALUATED FOR 10CFR50.59 APPLICABILITY: | None DESCRIBE WHERE THE ADDENDUM WILL BE EVALUATED FOR 10CFR50.59 APPLICABILITY: | ||
RAD/SCREEN/EVAL 13-02880 for Revision 1 of this remains applicable. | RAD/SCREEN/EVAL 13-02880 for Revision 1 of this remains applicable. | ||
LIST SUPPORTING DOCUMENTS: (Include total number of pages) Design Verlfjcatlon Record 1 page Calculation Review Cttecklist 3 pages Design Interface Summary 7 pages LIST ATTACHMENTS: (Include total number of pages) None | LIST SUPPORTING DOCUMENTS: (Include total number of pages) Design Verlfjcatlon Record 1 page Calculation Review Cttecklist 3 pages Design Interface Summary 7 pages LIST ATTACHMENTS: (Include total number of pages) None 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS Certain structures, components and systems of the nuclear plant are considered important to safety because they perform safety actions required to avoid or mitigate the consequences of abnormal operational transients or accidents. | ||
The purpose of this section is to classify structures, components and systems according to the importance of the safety function they perform. In addition, design requirements are placed upon such equipment to assure the proper performance of safety actions, when required. | |||
3.2.1 SEISMIC CLASSIFICATION Plant structures, systems and components important to safety are designed to withstand the effects of a Safe Shutdown Earthquake (SSE) and remain functional if they are necessary to assure: a. The integrity of the reactor coolant pressure boundary, b. The capability to shut down the reactor and maintain it in a safe condition, or c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of <10 CFR 100>. Plant structures, systems and components (including their foundations and supports) designed to designated as Structures, components, Class 1, Safety Class 2 or Safety discussion of safety classes) are except for (1) those noted in <Table 3.2-1> a or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>} 3.2-1 the event of an SSE are in <Table 3.2-1>. designated as Safety e <Section 3.2.3> for a Category I (2) those portions of Revision 12 January, 2003 No changes to this page. Provided for information. | |||
OF STRUCTURES, COMPONENTS AND SYSTEMS Certain structures, components and systems of the nuclear plant are considered important to safety because they perform safety actions required to avoid or mitigate the consequences of abnormal operational transients or accidents. | |||
The purpose of this section is to classify structures, components and systems according to the importance of the safety function they perform. In addition, design requirements are placed upon such equipment to assure the proper performance of safety actions, when required. | |||
CLASSIFICATION Plant structures, systems and components important to safety are designed to withstand the effects of a Safe Shutdown Earthquake (SSE) and remain functional if they are necessary to assure: a. The integrity of the reactor coolant pressure boundary, b. The capability to shut down the reactor and maintain it in a safe condition, or c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of <10 CFR 100>. Plant structures, systems and components (including their foundations and supports) designed to designated as Structures, components, Class 1, Safety Class 2 or Safety discussion of safety classes) are except for (1) those noted in <Table 3.2-1> a or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>} 3.2-1 the event of an SSE are in <Table 3.2-1>. designated as Safety e <Section 3.2.3> for a Category I (2) those portions of Revision 12 January, 2003 No changes to this page. Provided for information. | |||
!New Page ... I or B) is also deactivated and the electric heating coil in the charcoal filter train is automatically energized upon receipt of an emergency signal. The emergency recirculation system causes the supply air to be filtered through the charcoal filter train (M26-D001A, B) before being distributed to the control room. This system is idle during normal plant operation. | !New Page ... I or B) is also deactivated and the electric heating coil in the charcoal filter train is automatically energized upon receipt of an emergency signal. The emergency recirculation system causes the supply air to be filtered through the charcoal filter train (M26-D001A, B) before being distributed to the control room. This system is idle during normal plant operation. | ||
During periods of loss of offsite power, emergency power will be supplied by the standby diesel generators. | During periods of loss of offsite power, emergency power will be supplied by the standby diesel generators. | ||
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For the FHAES, the alternative source term FHA analysis took no credit for the charcoal adsorbers. | For the FHAES, the alternative source term FHA analysis took no credit for the charcoal adsorbers. | ||
The CRERS and FHAES charcoal adsorber beds are 2 inches deep. Exhaust air for both plenums is maintained at less than 70 percent relative humidity. | The CRERS and FHAES charcoal adsorber beds are 2 inches deep. Exhaust air for both plenums is maintained at less than 70 percent relative humidity. | ||
The HEPA filter efficiency of all the plenums is 99.97 percent on particles | The HEPA filter efficiency of all the plenums is 99.97 percent on particles 0.3 microns and larger. Has eal'teR fer fyel RaREl:liR§ aeeiel:eRe. | ||
and larger. Has eal'teR fer fyel RaREl:liR§ aeeiel:eRe. | |||
Additional bases for the design of the C RS, FHAES and AEGTS are presented in <Section 6.4>, <Section respectively. | Additional bases for the design of the C RS, FHAES and AEGTS are presented in <Section 6.4>, <Section respectively. | ||
6.5.1.2 System* Design 9.4.2>, and <Section 6.5.3> the LOCA analysis only credits the HEPA filters in the AEGTS and CRERS at an efficiency of 99 percent. The other design basis radiological calculations do not take credit for the t:tEPA filters In the AEGTS, CRERS, or FHAES. The design features of the CRERS, FHAES and AEGTS are compared to the recommendations of <Regulatory Guide 1.52> in <Table 6.5-1>, <Table 6.5-2>, and <Table 6.5-3> respectively. | 6.5.1.2 System* Design 9.4.2>, and <Section 6.5.3> the LOCA analysis only credits the HEPA filters in the AEGTS and CRERS at an efficiency of 99 percent. The other design basis radiological calculations do not take credit for the t:tEPA filters In the AEGTS, CRERS, or FHAES. The design features of the CRERS, FHAES and AEGTS are compared to the recommendations of <Regulatory Guide 1.52> in <Table 6.5-1>, <Table 6.5-2>, and <Table 6.5-3> respectively. | ||
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6.5-5 Revision 12 January, 2003 6.5.1.6 Materials No changes to this page. Provided for context. jNew Page ... I Estimated quantities of materials used in the activated charcoal adsorber plenums for CRERS, FHAES and AEGTS are listed in <Table 6.5-4>, <Table 6.5-5>, and <Table 6.5-6> respectively. | 6.5-5 Revision 12 January, 2003 6.5.1.6 Materials No changes to this page. Provided for context. jNew Page ... I Estimated quantities of materials used in the activated charcoal adsorber plenums for CRERS, FHAES and AEGTS are listed in <Table 6.5-4>, <Table 6.5-5>, and <Table 6.5-6> respectively. | ||
The governing specifications for the various materials are also listed and provide information regarding chemical composition of materials used. There are no radiolytic or pyrolytic decomposition products from the ESF filter systems. Actuation of the activated charcoal adsorber plenum water deluge fire protection systems will extinguish a charcoal fire before pyrolytic decomposition products are formed. None of these systems are located in areas where gamma radiation sources are sufficiently strong to cause radiolytic decomposition products. | The governing specifications for the various materials are also listed and provide information regarding chemical composition of materials used. There are no radiolytic or pyrolytic decomposition products from the ESF filter systems. Actuation of the activated charcoal adsorber plenum water deluge fire protection systems will extinguish a charcoal fire before pyrolytic decomposition products are formed. None of these systems are located in areas where gamma radiation sources are sufficiently strong to cause radiolytic decomposition products. | ||
Therefore, decomposition products do not affect any engineered safety features. | Therefore, decomposition products do not affect any engineered safety features. | ||
6.5.2 CONTAINMENT SPRAY SYSTEM 6.5.2.1 Design Bases a. The containment spray system (CSS) is a part of the residual heat removal (RHR) system. | |||
SPRAY SYSTEM 6.5.2.1 Design Bases a. The containment spray system (CSS) is a part of the residual heat removal (RHR) system. | |||
* b. The CSS provides containment cooling following a loss-of-coolant accident, in addition to being a fission product removal mechanism. | * b. The CSS provides containment cooling following a loss-of-coolant accident, in addition to being a fission product removal mechanism. | ||
Refer to <Section 6.2.2> for the heat removal function of the CSS. c. The CSS consists of two completely redundant and independent loops. d. The CSS is designed to remain operable in the containment accident environment, which is discussed in <Section 3.11>. 6.5-9 Revision 12 January, 2003 INew Page ... I are not subject to clogging by particles less than 1/4 inch in maximum dimension. | Refer to <Section 6.2.2> for the heat removal function of the CSS. c. The CSS consists of two completely redundant and independent loops. d. The CSS is designed to remain operable in the containment accident environment, which is discussed in <Section 3.11>. 6.5-9 Revision 12 January, 2003 INew Page ... I are not subject to clogging by particles less than 1/4 inch in maximum dimension. |
Revision as of 11:01, 11 May 2019
ML14203A625 | |
Person / Time | |
---|---|
Site: | Perry |
Issue date: | 07/22/2014 |
From: | Harkness E J FirstEnergy Nuclear Operating Co |
To: | Document Control Desk, Office of Nuclear Reactor Regulation |
References | |
L-14-245, TAC MF3197 | |
Download: ML14203A625 (93) | |
Text
FENOCŽ RrstEnergy Nuclear Operating Company Em8st J. Harkness Vice President July 22, 2014 L--*14-245 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Perry Nuclear Power Plant P.O. Box97 10 Center Road Perry. Ohio 44081 440-280-5382 Fax: 440-280-8029 10 CFR 50.90 Response to Request for Additional Information Regarding Request for Licensing Action on Alternative Accident Source Term Radiological Dose Calculations (TAC No. MF3197) A request for licensing action regarding alternative accident source term radiological dose calculations was submitted to the Nuclear Regulatory Commission (NRC) by letter dated December 6, 2013 (Accession No. ML 13343A013).
The NRC staff requested additional information in a letter dated June 24, 2014 (Accession No. ML 14162A409).
The requested information is provided in Attachment
- 1. In addition, the effects of several necessary changes to the supporting dose calculations are presented in Attachment
- 2. No changes were identified to the previously provided Significant Hazards Consideration.
There are no regulatory commitments contained in this submittal.
If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager-Fleet Licensing, at 330-315-6810.
I declare under penalty of perjury that the foregoing is true and correct. Executed on July 2Z. , 2014. Sincerely, Ernest J. Harkness Attachments:
- 1. Response to Request for Additional Information on Radiological Dose Calculations
- 2. Effects of Necessary Changes to Supporting Radiological Dose Calculations cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager State of Ohio (NRC Liaison) Utility Radiological Safety Board "\'.,
Attachment 1 L-14-245 Response to Request for Additional Information on Radiological Dose Calculations Page 1 of 16 FirstEnergy Nuclear Operating Company (FENOC) provided a request for licensing action (RLA) regarding alternative accident source term radiological dose calculations to the Nuclear Regulatory Commission (NRC) in a letter dated December 6, 2013. The NRC staff requested additional information in a letter dated June 24, 2014 .. The requested information is identified using bold text, followed by the FENOC response.
- 1. For those changes to the current licensing basis (CLB) parameters used in the affected dose consequence analyses, provide additional Information describing for each affected design basis accident, all the basic parameters used in the dose consequence analyses.
For each parameter, please indicate the CLB . value, the revised value where applicable, and the basis for any. changes made to the CLB values. The U.S. Nuclear Regulatory Commission (NRC) staff requests that this information be presented in separate tables for each accident evaluated.
Response:
A matrix is provided for the loss of coolant accident (LOCA), the control rod drop accident (CRDA) scenarios, and the main steam line break outside containment (MSLBOC) analyses.
LOCA Parameter Current Licensing Basis Proposed Licensing Basis for changes [Existing location]
Basis [Proposed location]
Core Source Term General Electric (GE)12/ Global Nuclear GNF2 fuel design to Basis GE14 Fuel (GNF)2 be introduced at [GE12= Amendment 112 [Updated Safety Analysis PNPP during next (power uprate), Report (USAR) fuel cycle NRC Safety Evaluation pg. 15.6-65] (SE) pg. 4; GE14= 10 CFR 50.59 review when converted from GE12] Power Level 102% of rated thermal 102% of rated thermal No change power = 3758 megawatts power = 3833 MWt (existing LOCA thermal (MWt) x 1.02 = [USAR pg. 15.6-24 & dose calculation in 3833 MWt 15.6-58 thru 60] support of [Amendment 112; NRC SE Amendment 112 did pg. 45] assume 3833 MWt although several USAR pages simply reflect the licensed 100% RTP value of 3758 MWt)
Attachment 1 L-14-245 Page 2of16 Isotopes in source term Dose Conversion Factors Gap Release Timing Credit for decay during 2 minutes before start of the gap release. Suppression Pool Scrubbing Drywell Volume Containment Volume (excluding drywell} Volume of Sprayed Region Volume of Containment Unsprayed Region Volume of Total Unsprayed Regions (including drywall} 76 isotopes [Calculations supporting Amendment 1 a3 (the PNPP pilot plant AST amendment}]
Federal Guidance Report (FGR) 11 [USAR pg. 15.a-37) 3a seconds [This assumption was not specified in the USAR but was assumed in the calculations supporting Amendment 1 a3] Not considered No Credit [USAR 15.6-27) 2. 765x1 as feet (ft}3 [USAR pg. 15.6-58] 1.1654x1 as ft 3 [USAR pg. 6.5-56 & 15.6-58) 4.81x1as ft 3 [USAR pg. 15.6-59) 6.84x1as ft 3 [USAR pg. 15.6-59) 9.6a7x1 as ft 3 [not currently specified separately in the USAR] 6a isotopes Based on isotopes [Request for licensing used in the action (RLA} submittal RADTRAD dated 12/6/2a13, and computer code Amendment TBD dated TBD] FGR 11and12 Use of Regulatory
[USAR pg. 15.a-37) Guide (RG} 1.183, Section 4.1.2 and 4.1.4. 2 minutes (BWR) Per RG 1.183, [USAR pg. 15.6-22 is Table4; revised to reference BWR-specific RG 1.183 for timing of the releases, and Table 4 of RG 1.183 specifies 2 minutes for onset of BWR gap release] Decay considered Utility decision since [USAR pg. 15.6-23) RG 1.183 does not specify this No Credit No change [USAR 15.6-27) 2. 765x1 as ft 3 No change [USAR pg. 15.6-58) 1.1654x10 6 ft 3 No change [USAR pg. 6.5-56 & (Note: Table 6.5-9 & 15.6-58) Table 15.6-12a markups are correcting typos) 4.812x1as ft 3 Essentially no [USAR pg. 15.6-59) change (third decimal point} 6.842x1as ft 3 Essentially no [USAR pg. 15.6-59) change (third decimal point} 9.6a7x1 os ft 3 No change [USAR pg. 15.6-59)
Attachment 1 L-14-245 Page 3of16 Flow Rate from Drywall to Unsprayed Region of the Containment Flow Rate from Unsprayed Region of the containment back to Drywell Primary Containment Leakage Rate (sprayed and unsprayed regions leakage) Percent of Primary Containment Leakage that goes Into the annulus Percent of Primary Containment Leakage that bypasses the Secondary Containment Credit for Reduction in Primary Containment Leakage based on containment pressure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0-2 hours (hrs.) 3000 cubic 0 -0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (hr.) 0 cfm feet per minute 0.5-2 hrs. 3000 cfm (cfm) 2 -720 hrs. 2.77x10 5 cfm 2_720 hrs. 2_7 , 7 x 105 cfm [USAR Table 15.6-12b on [USAR Table 15.6-12b on pg. 15*6-59 1 pg. 15.6-59] 0 -2 hrs. 0 cfm 2 -720 hrs. 2. 77x10 5 cfm [USAR pg. 15.6-59] 0.2 percent of containment atmosphere per day (La) [USAR pg. 6.2-72 & 73; Technical Specification (TS) Section 5.5.12] 89.92 percent (0.8992 La) [USAR 15.6-31] 10.08 percent (0.1008 La) [USAR pg. 15.6-31] Not taken 0 -2 hrs. O cfm 2 -720 hrs. 2.77x10 5 cfm [USAR pg. 15.6-59] 0.2 percent of containment atmosphere per day (La) [USAR pg. 6.2-72 & 73; TS Section 5.5.12] 89.92 percent (0.8992 La) [USAR 15.6-31] 10.08 percent (0.1008 La) [USAR pg. 15.6-23 & 31) Based on containment pressure reduction after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, leakage is reduced to 69 percent of the value used during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [USAR pg. 15.6-30] No actual change (Note: USAR Table 15.6-12b did not reflect the zero flow rate out of the drywall during the first 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period that was in the existing calculation)
No actual change (USAR markup of page 15.6-59 simply rounds 2. 765 to 2.77 x 10 5 cfm, which is the value in the existing calculation)
No change No change (Rev. 0 of new LOCA calculation used 100 percent, but Rev. 1 corrects that to 89.92 percent) No change Credit taken for reduction in containment pressure, as permitted In RG 1.183. See pages 29 & 30 of the RLA Degree of Conformance matrix Attachment 1 L-14-245 Page 4of16 Credit for Reduction In Secondary Containment Bypass Leakage based on containment pressure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Containment Spray Initiation Containment Spray Duration Spray Fall Height Containment Spray-Aerosol (particulate)
Removal Other Inputs to Spray Modeling Not taken 1 O minute auto initiation (based on high pressure and LOCA signal) [USAR 15.6-32] 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [USAR pg. 6.5-12 & 15.6-32] 53.2 feet [USAR pg. 15.6-59) Sprayed region of the containment modelled using STARNAUA methodology
[USAR pg. 6.5-13, 14, 56, & 58) (Note: In Am. 103, NRC used RADTRAD goth percentile uncertainty distributions, per pg. 10 of the NRC SE for Am. 103) Various (See USAR Table 6.5-9 "Input Parameters For The Spray Removal Analysis" [USAR pg. 6.5-56] After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, leakage is Credit was taken for reduced to 69 percent of reduction in the value used during the containment first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> pressure, as (USAR pg. 15.6-30) permitted in RG 1.183 30 minute manual initiation Initiation time is [USAR 15.6-32) conservatively tripled, because. auto-initiation on containment high pressure might not be achieved 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> No change [USAR pg. 6.5-12 & 15.6-32] 54.05 feet New calculation
[USAR pg. 15.6-59) Sprayed region of the Conservative containment modelled RADTRAD model using RADTRAD Powers assumption, model with 10 1 h percentile RG 1.183, uncertainty distribution.
Appendix A, Particulate removal Section 3.3 coefficient due to sprays is reduced by a factor of 10 after the aerosol mass is depleted by a factor of 50 [USAR pg. 6.5-13, 14, & 56) Q = 0.0621 (Spray Flux) Powers spray model Alpha = 1.422 (Ratio of inputs unsprayed volume to sprayed volume) [USAR pg. 6.5-56)
Attachment 1 L-14-245 Page 5of16 Containment Spray-Elemental Iodine Removal Natural Deposition-Aerosol (particulate)
Removal Natural Deposition-Elemental Iodine Removal Engineered Safety Feature (ESF) Leakage Pathway Direct to Environment Offsite Breathing Rates (meters (m)3/ second (sec)) Sprayed region of the containment modelled using STARNAUA methodology
[USAR pg. 6.5-13, 14, 56, & 58) Drywall region modelled using STARNAUA methodology
[USAR pg. 15.6-28 & 29) (Note: In Am. 103, NRC used RADTRAD 90lh percentile uncertainty distributions, per pg. 7 of the NRC SE for Am. 103) Drywall region modelled using STARNAUA methodology (USAR pg. 15.6-30) Time (hours) Leak Rate 0-24 15 gallons per hour (gph) 24-24.5 15 gph + 50gpm 24.5-720 15gph [USAR pg. 15.6-33 & 60] Time (hours) 0-8 3.47x10 4 8-24 1.75x10 4 24-720 2.32x10 4 [USAR pg. 15.0-37) Sprayed region of the RG 1.183 and containment modelled SRP6.5.2 using Standard Review Plan (SRP) 6.5.2 guidance.
Elemental Iodine removal by sprays is terminated when a OF of up to 200 is reached [USAR pg. 6.5-13, 14, & 56) Drywall and unsprayed Conservative volume of containment RADTRAD "'odel modelled using RADTRAD assumption Powers Model for aerosol removal with 10 1 h percentile uncertainty distribution
[USAR 15.6-28 & 29) Drywall, sprayed, and RG 1.183 and unsprayed volume of SRP6.5.2 containment modelled using SRP 6.5.2 guidance.
Similar to the spray assumptions, elemental Iodine removal is terminated when a OF of up to 200 is reached [USAR pg. 6.5-13, & Table 6.5-11) Time (hours) Leak Rate No change 0 .. 24 15gph 24-24.5 15 gph + 50gpm 24.5-720 15gph [USAR pg. 15.6-23, 33 & 60] Time (hours) RG 1.183, 0-8 3.5x10 4 Section 4.1.3 8""'.'24 1.8x10 4 24-720 2.3x10 4 [USAR pg. 15.0-37)
Attachment 1 L-14-245 Page 6of16 Main Steam Line (MSL) leak rate (standard conditions)
! Main Steam Line leak rate converted from standard to non-standard conditions (assuming elevated post-accident temperatures) during first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Credit for Main Steam line leakage reduction associated with decreased pressure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 250 standard cubic feet per hour (scfh) total, 100 scfh maximum per line (under standard conditions).
Modelled as 100 scfh to the broken MSL, and 150 scfh to the intact MSL's [TS Surveillance Requirement (SR) 3.6.1.3.1 O] Flow rate from drywall to all steam lines (both broken and intact) = 298 cfh for first 7 484 seconds, and 247 cfh until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Flow rate between MSIVs in one steam line = 191 cfh (Note: USAR Table 15.6-12a says cfm vs. cfh, which was a typographical error) [USAR pg. 15.6-58] Not taken [USAR 15.6-25 & 58] 250 scfh total, 100 scfh No change maximum per line (under standard conditions).
Modelled as 100 scfh to the broken MSL, and 150 scfh to the intact MSL's [TS SR 3.6.1.3.1 O] Flow rate from drywall to No change the broken steam line (values in USAR = 1. 987 cfm for first 7 484 Table 15.6-12a seconds, and 1.647 cfm appear to be until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> changed, but that is Flow rate from drywell to due to mathematical the intact steam lines conversion of . = 2.98 cfm for first 7484 existing assumed seconds, and 2.47 cfm flow rates from cfh until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to cfm (engineering Flow rate in one steam unit change only), line, between the MSIVs and inclusion of = 3.183 cfm more detail into this Flow rate in the other (the table about how intact) steam lines, flow rates to (and between the MSIVs through) the main =4.775 cfm steam lines are [USAR pg. 15.6-58] attributed to the broken versus the intact main steam lines) After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, leakage Some credit was from the drywell to the taken for reduction Main Steam Lines is in containment reduced to 69 percent of pressure, as the value used during the permitted in first 24 hrs. However, the RG 1.183 leak rate assumed between the MSIVs is conservatively not reduced after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . [USAR pg. 15.6-25 & 58]
Attachment 1 L-14-245 Page 7of16 Aerosol Removal Efficiencies in Steam Lines Annulus Exhaust Gas Treatment (AEGT) system recirculation to the annulus AEGT Filtration (High Efficiency Particulate Air (HEPA) filter) Broken steam line -between MSIVs (100 scfh) -For Particulate Iodine, solubles, and insoluble 68.1 % from 0 to 0.5 hr. 83.5% from 0.5 to 1.5 hrs. 87.13% from 1.5 to 3 hrs. 89% from 3 to 5 hrs. 86.14% from 5 to 7 hrs. 81.85% from 7 to 9 hrs. 76.9% from 9 to 11 hrs. 36.53% from 11 to 720 hrs. -For Elemental Iodine -45% Intact Steam lines RPV to IB MSIV (150 scfh) -For Particulate Iodine, solubles and insoluble 72.06% from 0 -end -For Elemental Iodine -45% Intact Steam Lines between MSIVs -For Particulate Iodine, solubles, and insoluble 71.4% from Oto 0.5 hr. 81.3% from 0.5 to 1.5 hrs. 83.61 % from 1.5 to 3 hrs. 84.49% from 3 to 5 hrs. 83.39% from 5 to 7 hrs. 79.98% from 7 to 9 hrs. 75.58% from 9 to 11 hrs. 38.07% from 11 to 720 hrs. -For Elemental Iodine -45% [USAR pg. 15.6-26 & 27 provide general discussion of steam line treatment]
Not considered, such that flow rate to the environment
= 2000 cfm [Not currently specified in the USAR] 99 percent efficiency
[USAR pg. 15.6-31 & 59] Broken steam line -No change between MSIVs (100 scfh) . -For Particulate Iodine, solubles, and insoluble 68.1 % from 0 to 0.5 hr. 83.5% from 0.5 to 1.5 hrs. 87.13% from 1.5 to 3 hrs. 89% from 3 to 5 hrs. 86.14% from 5 to 7 hrs. 81.85% from 7 to 9 hrs. 76.9% from 9 to 11 hrs. 36.53% from 11 to 720 hrs. -For Elemental Iodine -45% Intact Steam lines RPV to 18 MSIV (150 scfh) -For Particulate Iodine, solubles and insoluble 72.06% from 0 -end -For Elemental Iodine -45% Intact Steam Lines between MSIVs -For Particulate Iodine, solubles, and insoluble 71.4% from 0 to 0.5 hr. 81.3% from 0.5 to 1.5 hrs. 83.61 % from 1.5 to 3 hrs. 84.49% from 3 to 5 hrs. 83.39% from 5 to 7 hrs. 79.98% from 7 to 9 hrs. 75.58% from 9 to 11 hrs. 38.07% from 11 to 720 hrs. -For Elemental Iodine -45% [USAR pg. 15.6-26 & 27 provide general discussion]
Not considered, such that No change flow rate to the environment
= 2000 cfm [USAR pg. 15.6-31] 99 percent efficiency No change [USAR pg. 6.5-4; 15.6-31 & 59) ,-.. : .*,'"
Attachment 1 L-14-245 Page 8of16 AEGT Filtration (Charcoal)
Exclusion Area Boundary (EAB) X/Q ( sec/m 3) Low Population Zone (LPZ) X/Q (sec/m 3) Control Room X/Q (sec/m 3) Control Room Volume Control Room Filtration System (HEPA) Control Room Filtration System (Charcoal)
Zero percent efficiency
[USAR pg. 6.5-3) 4.3x10-4 [NRC Safety Evaluation for Amendment 103, pg. 15) Time (hrs) 0-8 4.8x1Q*S 8-24 3.3x1Q*S 24-96 1.4x1Q*S96-720 4.1x10.a [NRC Safety Evaluation for Amendment 103, pg. 15) Time (hrs) 0-8 3.5x10-4 8-24 2.1x10-4 24-96 1.1x10-4 96-720 5.75x10-5 [NRC Safety Evaluation for Amendment 103, pg. 14) 3.44x10 5 ft 3 [USAR pg. 15.6-64) 95 percent credit for HEPA removal of Particulates.
[USAR pg. 15.6-35) 50 percent charcoal filter efficiency for elemental and organic iodine. [USAR pg. 6.5-4, 15.6-23 & 35) Zero percent efficiency No change [USAR pg. 6.5-3 & 42) 4.3x10-4 No change [USAR pg. 15.6-63) Time (hrs) No change 0-8 4.8x10-5 8-24 3.3x10-5 24-96 1.4x10-5 96-720 4.1x10.a [USAR pg. 15.6-63) Time (hrs) No change 0-8 3.5x10-4 (markups on USAR 8-24 2.1x10-4 page 15.6-63 are 24-96 1.1x10-4 correcting incorrect 96-720 5.75x10-5 information)
[USAR pg. 15.6-63 with corrections to reflect correct CLB] 3.90x10 5 ft 3 Calculation for [USAR pg. 15.6-64) control room volume revised since pilot plant submittal 99 percent credit for HEPA Acceptable value removal of Particulates based on current [USAR pg. 6.5-4; 15.6-35) and future TS test acceptance criterion 80 percent charcoal filter Eliminated an efficiency for elemental unnecessary and organic iodine. conservatism;
[USAR pg. 6.5-4 & 30, 80 percent value will 15.6-23 & 35) support a future request for licensing action to revise TS charcoal adsorber testing acceptance criterion to a value of 10% penetration
(=90% efficiency) versus the current 2.5% penetration requirement Attachment 1 L-14-245-Page 9of16 Control Room Emergency Recirculation System ActuationfTiming of the Start of Filtration Timing of the Reduction of Unfiltered lnleakage Into the Control .Room Control Room Breathing Rate Control Room Occupancy Factors 30 minutes [USAR pg. 6.4-8, 15.6-23) 1375 cfm unfiltered inleakage starting at time (t)=O, for duration [USAR pg. 6.4-14 & 15, 15.6-35 & 64) 3.47x10-4 m 3/sec [USAR pg. 15.0-37) Time (hours) 0-24 1.0 24-96 0.6 96-720 0.4 [not currently specified in the USAR] 30 minute manual operator No credit taken for actuation automatic actuation of filtration or for [USAR pg. 6.4-8, 14, & 15, any reduction in 15.6-23 & 35 unfiltered In-leakage until 30 minutes, to support a future request for licensing action to remove TS controls over auto-initiation instrumentation 6600 cfm unfiltered Conservatively inleakage from t=O to assumes outside air 30 minutes inlet is not isolated 1375 cfm from t=30 min to until t=30 minutes when Control Room 30 days Emergency
[USAR page 6.4-14 & 15, Recirculation 15.6-35 & 64) System is manually actuated, to support a future request for licensing action to remove TS controls over auto-initiation instrumentation 3.5x10-4 m 3/sec RG 1.183, [USAR pg. 15.0-37) Section 4.2.6 Time (hours) No change 0-24 1.0 24-96 0.6 96-720 0.4 [USAR pg. 15.6-64) CRDA (Scenario 1 -Standard Review Plan Section 15.4.9) Parameter Current Licensing Basis Proposed Licensing Basis Basis For , [Current Location]
[Proposed Location]
Changes Core Source Term GE14 GNF2 Change in basis [USAR pg. 15.4-27, 45, & [USAR pg. 15.4-27, 45, & 46) fuel design 46]
Attachment 1 L-14-245 Page 10of16 Power Level Coincident loss of offsite power (LOOP)? Isotopes Considered Assumed number of failed fuel rods Iodine Fractions, % Radial Peaking Factor Gap and Melt fraction Activity released to condenser Activity available for release from condenser Plant startup, at low power level, after operation at 3,833MW1 [USAR pg. 15.4-26 thru 29, &44) Yes [USAR pg. 15.4-27 & 28] Considered Iodine, Krypton, and Xenon isotopes [USAR pg. 15.4-29] 1107 [USAR pg. 15.4-27, 28, & 44] Organic 0 Elemental 100 Particulate 0 [USAR pg. 15.4-44] 1.7 [Reload Analysis Parameter]
Group Gap Melt Noble Gas 10% 100% Halogen 10% 50% [USAR pg. 15.4-29] Group Noble Gas 100% Halogen 10% [USAR pg. 15.4-29] Group Noble Gas 100% Halogen 10% [USAR pg. 15.4-29] Plant startup, at low power level, No change after operation at 3,833 MW1 [USAR pg. 15.4-26 thru 29, & 44) Yes No change [USAR pg. 15.4-27 & 28] (per SRP 15.4.9 Section Ill, Assumption
- 1) Considered Iodine, Krypton, Added Xenon, Bromine, Cesium, and Isotopes per Rubidium isotopes RG 1.183, [USAR pg. 15.4-29] Section 3, Tables 3 & 5 1376 (two* full rows of bundles Conservative around the dropped control rod) increase In number of [USAR pg. 15.4-27, 28, & 44] failed rods Organic 0.15 RG 1.183, Elemental 4.85 Appendix C, Particulate 95.0 Item 3.6 [USAR pg. 15.4-44] 2.0 Current radial [Reload Analysis Parameter]
peaking factor in analyses Group Gap Melt Added alkali Noble Gas 10% 100% metals per Halogen 10% 50% RG 1.183, Alkali Metals 12% 25% Section 3.2, [USAR pg. 15.4-29] Tables 1 & 3 Group RG 1.183, Noble Gas 100% Appendix C, Halogen 10% Item 3.3 Remaining 1% [USAR pg. 15.4-29] Group RG 1.183, Noble Gas 100% Appendix C, Halogen 10% Item 3.4 Particulate 1% [USAR pg. 15.4-29]
Attachment 1 L-14-245 Page 11of16 Condenser Leakage Release to Environment Atmospheric Dispersion X/Q The leakage rate from the condenser is assumed to be 1 % per day for the duration of the event (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). [USAR pg. 15.4-29 (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> termination time is from SRP 15.4.9 Assumption 111.12)] All activity leaking from the condenser is assumed to leak directly to the environment without mixing in the turbine building.
[USAR pg. 15.4-28 & 44] Time EAB LPZ 0-2 hrs. 4.3E-4 4.SE-5 2-8 hrs. --4.SE-5 8-24 hrs. --3.3E-5 1-4 Days --1.4E-5 4-30 Days --4.1 E-6 [USAR pg. 15.4-45] Control Room Dose Not calculated Control Room Emergency Recirculation System Filtration (Charcoal and HEPA) AEGT System Filtration (Charcoal and HEPA) [Control room doses were not previously required to be determined for CRDA] NIA [Control room doses were not previously required to be determined for CRDA] Not
[Releases are from outside containment, so AEGT system is N/A, therefore it is not credited]
The leakage rate from the condenser is assumed to be 1 % per day for the duration of the event (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) [USAR pg. 15.4-29] All activity leaking from the condenser is assumed to leak directly to the environment without mixing in the turbine building.
[USAR pg. 15.4-28 & 44] Time EAB LPZ CR 0-2 hrs. 4.3E-4 4.8E-5 3.5E-4 2-8 hrs. --4.8E-5 3.5E-4 8-24 hrs. --3.3E-5 2.1 E-4 1-4 Days --1.4E-5 1.1 E-4 4-30 Days --4.1 E-6 5. 75E-5 [USAR pg. 15.4-45] Calculated
[USAR pg. 15.4-47] Not credited [USAR pg. 6.5-4 & 30, 15.4-45] Not credited [USAR pg. 6.5-3, 4, & 42] No change No change No change except for the addition of control room (CR) X/Q's RG 1.183, Section 4.4 No credit taken No change Attachment 1 L-14-245 Page 12of16 CRDA (Scenario 2 -NED0-31400 11 Safety Evaluation for Eliminating the Bolllng Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor")
Parameter Current Licensing Basis Proposed Licensing Basis Basis For {See Scenario 1, [Current Location]
[Proposed Location]
Changes with the following changes based on NED0-31400}
Power Level Plant startup, at low power Plant startup, at low power level, No change level, after operation at after operation at 3,833 MW1, 3,833 MW1, aligned to the aligned to the Offgas system Offgas system [USAR pg. 15.4-26, 27, 28, & 30) [USAR pg. 15.4-26, 27, 28, & 30) Coincident loss of No, so Offgas System No, so Offgas System remains No change offsite power remains running to hold up running to hold up all but the (LOOP)? all but the noble gases noble gases [USAR 15.4-30) [USAR 15.4-30) MSLBOC Current Licensing Basis Proposed Licensing Basis Basis for Parameter
[Current Location]
[Proposed Location]
Changes Reactor Coolant Original GE-supplied Revised GE-supplied reactor RG 1.183, Inventory reactor coolant inventory coolant inventory for: Appendix D, for: a) Design basis analysis = TS pre-Assumption 2 a) Design basis analysis = accident spike of 4.0 µCi/gm, and TS pre-accident spike of b) Realistic analysis = 4.0 microcuries
(µCi)/ TS maximum equilibrium value of gram (gm), and 0.2 µCi/gm b) Realistic analysis = [USAR pg. 15.6-13 & 15) TS maximum equilibrium value of 0.2 µCi/gm [USAR pg. 15.6-13 & 15] Source Term Considered Iodine, Krypton, Considered Iodine, Krypton, Revised and Xenon isotopes as Xenon, and Bromine isotopes as GE-supplied coolant activity coolant activity reactor [USAR pg. 15.6-52 & 56) [USAR pg. 15.6-52 & 56) coolant isotopes Iodine Fractions Organic 0 Organic 0.15 RG 1.183, (%) Elemental 100 Elemental 4.85 Appendix D, Particulate 0 Particulate 95.0 Assumption
[USAR pg. 15.6-54] [USAR pg. 15.6-54) 4.4 Attachment 1 L-14-245 Page 13of16 Isolation Valve Closure Time, and mass of coolant released to the environment Atmospheric Dispersion X/Q Control Room Dose Control Room Emergency Recirculation System Filtration (Charcoal
& HEPA) AEGT System Filtration (Charcoal and HEPA) Various closure times are listed on current USAR pages, but the resultant mass release (consisting of 14,311 pounds.of steam and 127 ,376 pounds of liquid) listed in the USAR is consistent with the supporting calculation
[USAR 15.6-12 & 13] EAB LPZ 6. 7E-4 8.2E-5 [USAR pg. 15.6-55] Not calculated
[Control room doses were not previously required to be determined for MSLBOC] N/A [Control room doses were not previously required to be determined for MSLBOC] Not credited [Releases are from outside containment, so AEGT system is NIA, therefore it is not credited]
The supporting calculation's.
conservative valve closure time of 6.05 seconds will be reflected in the USAR, with its resultant mass release of 14,311 pounds of steam and 127,376 pounds of liquid [USAR 15.6-12 & 13] Time EAB LPZ CR 0-2 hrs. 4.3E-4 4.8E-5 3.5E-4 2-8 hrs. --4.8E-5 3.5E-4 8-24 hrs. -3.3E-5 2.1 E-4 1-4 Days -1.4E-5 1.1E-4 4-30 Days -4.1 E-6 5. 75E-5 [USAR pg. 2.3-89 & 15.6-55] Calculated
[USAR pg. 15.6-53, 55 & 57) Not credited [USAR pg. 6.5-4 & 30) Not credited [USAR pg. 6.5-3, 4, & 42) Clarification of actual (conservative) valve closure timing in the existing calculation; no change in the amount of reactor coolant released Utilized updated X/Q values and added control room RG 1.183, Section 4.4 No credit taken No change Attachment 1 L-14-245 Page 14of16 2. On page 7 of the submittal evaluation, FirstEnergy Nuclear Operating Company addressed the impact of the proposed*
changes on the radiological habitability of the Technical Support Center. However, the application appears to be silent on the impact of those changes on the calculations supporting the establishment of numeric thresholds for emergency action levels (EALs) related to readings on: (1) radiological effluent radiation monitors, and (2) the in-containment high range radiation monitor readings used in the fission product barrier matrix EALs. In addition to the source term change and core inventory changes, the NRC staff notes that you have proposed changes in other analysis assumptions (e.g., filter removal efficiencies, containment leakage rates, drywell flows, containment spray credit, etc.) that could potentially affect the validity of the previous EAL thresholds.
Explain the impact of the changes on these EAL thresholds.
Response:
The current PNPP Emergency Plan utilizes the Nuclear Utility Management and Resources Council (NUMARC) EAL methodology, NUMARC/NESP-007, Methodology for Development of Emergency Action Levels. When the plant was originally licensed in the mid-1980's, the dose assessment program utilized a source term based on the licensing-basis loss of coolant accident (LOCA) in the Final Safety Analysis Report (FSAR). Subsequently, in October of 1988, the NRC published NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents.
This guidance reflected a more current understanding of accident source terms and their offsite consequences.
The NUREG noted on page 1-6 that licensing basis evaluations
."normally should not be used to estimate actual accident source terms or offsite doses." The NUREG (beginning on page 2-5) provided core inventory
- assumptions to be used for source term estimation, expressed in curies per electrical megawatt (Ci/MWe), such that the source term could be adjusted for various plants. The NUREG explained on page 1-2 that "studies of the uncertainties associated with source term estimation indicate that source term projections based on accident conditions are only accurate within a factor of 100 or more ... " Therefore, the generic, scalable source term provided by NUREG-1228 is considered to provide an acceptable standard for use at various plants, including PNPP. In the early 1990s at PNPP, when dose assessment software was updated, the source term provided in NUREG-1228, scaled for PNPP's licensed power level at that time, was utilized.
- The dose assessment program was updated again in 2012, utilizing the Meteorological Information and Dose Assessment Software (MIDAS) program. As noted in NUREG-1228 on page 1-3, "if a change in assumptions does not result in a change to the source term by at least 1 order of magnitude, it is not worth considering
... " Therefore, it was determined that the NUREG-1228 source term, scaled for PNPP's five percent uprate, would be used as input for the dose assessment program (for events postulated to occur with the plant at power). The.dose assessment program was utilized to calculate plant effluent monitor values that would result in 1 Rem Total Effective Dose Equivalent (TEDE) and/or 5 Rem Child Thyroid doses offsite (the current EAL thresholds).
Similarly, for item (2) in the NRC RAI (the in-containment high-range radiation monitors), the PNPP EALs are based on the NUMARC/NESP-007 guidance.
That guidance Attachment 1 L-14-245 Page 15of16 discusses the determination of in-containment radiation monitor readings for fuel failure events, specifically, for the fuel clad barrier on page 5-20 and the primary containment barrier on page 5-23.
5-23 includes a reference to NUREG-1228.
Similar to the discussions above for the effluent radiation monitors, the NUREG-1228 conclusion that unless a change in assumptions results "in a change to the source term by at least .1 order of magnitude, it is not worth considering" can also be applied to the In-containment radiation monitor reading thresholds.
Since the licensing basis source term is not changing significantly with the adoption of the GNF2 fuel, and a NUREG-1228-based source term based on curies per electrical megawatt would not change, it is concluded that the PNPP in-containment radiation monitor reading values specified in the EALs do not need to be changed as a result of the proposed change to the licensing basis source term for the LOCA event. In summary, it is concluded that implementation of a new licensing basis source term, such as proposed by this request for licensing action (RLA), does not invalidate the current numeric thresholds in the Emergency Plan EALs, including the (1) radiological effluent radiation monitors, and (2) the in-containment high range radiation monitors.
Other proposed changes to licensing basis analysis assumptions, such as those listed in the NRC RAI (filter efficiencies, containment leakage rates, drywell flows, containment spray credit, etc.), also would not invalidate use of the current numeric thresholds in the EALs. 3. Address whether any non-safety-related systems and components are credited in the accident source term (AST) analyses.
If so: a. Describe the independence (electrical and physical separation) of these systems from the safety-related systems. Provide a detailed discussion on why a fault on the non-Class 1 E electrical circuit will not propagate to the Class 1 E electrical circuit. b. Describe the redundancy of these systems and how these systems meet the single failure criterion.
Response:
No non-safety-related systems or components are credited to reduce doses in the alternative AST analyses.
Conservative assumptions in several event dose analyses assume the normal control room ventilation system continues to operate for various lengths of time post-accident, which increases the amount of unfiltered inleakage into the control room versus use of an assumption that the systems shut down or otherwise isolate quickly, so this is not considered to be a 'credit.'
The Scenario 1 CRDA analysis notes that if the mechanical vacuum pumps are running at a very low plant power level, the pumps will shut down as a result of the Standard Review Plan 15.4.9-required Assumption 1 of a loss of offsite power (LOOP); this has been previously reviewed and approved by the NRC for the CRDA at PNPP.
Attachment 1 L-14-245 Page 16of16 4. Address whether any loads are being added to the emergency diesel generators (EDGs). If so, describe their Impact on the capability and capacity of the EDGs. Also, describe changes, if any, being made to the EOG loading sequence to support this license amendment request. Response:
No new loads are being added to the EDGs, and there are no changes being made to the EDG loading sequence in support of this amendment request. 5. Discuss why there are no changes to the Environmental Qualification profiles as a result of the full implementation of the AST. Response:
A calculation was performed to determine the impact on equipment environmental qualification due to the post-accident fuel failure radiation dose for GNF2 fuel. The maximum radiation dose rate and the integrated dose (over the course of an . accident) following failure of GNF2 fuel was compared to the maximum dose rate and the integrated dose for the current design basis source term. It was determined that the accident doses from the GNF2 source term trended closely with the original source term, with the GNF2 fuel post-accident doses being slightly higher. The increases for the maximum dose rates were on the order of 10.5 percent, and the integrated dose increases ranged from zero percent to approximately eight percent. The increases represent the expected change in the post-accident dose profiles as a result of the transition to the GNF2 fuel. Equipment Qualification packages were reviewed to determine if these increases would exceed the existing radiation qualification of equipment.
In all cases, the existing qualification doses bound the expected increased doses resulting from the transition to GNF2 fuel. Formal updates to the environmental
- qualification zone profiles, equipment calculations, and affected auditable file packages, are ongoing to reflect the results of the above review of the GNF2 post-accident doses.
Attachment 2 L-14-245 Effects of Necessary Changes to Supporting Radiological Dose Calculations Page 1of76 Several necessary changes were identified in two of the calculation summaries included as addendums to the request for licensing action (RLA) submitted on December 6, 2013; specifically Addendum 4, "Summary of Loss of Coolant Accident (LOCA) Dose Calculation," and Addendum 5, "Summary of Control Rod Drop Accident (CRDA) Dose Calculation." Updated summaries of the calculations are included in this attachment, following this discussion (42 pages in the LOCA summary, and 2 pages in the CRDA summary).
The issue of these calculations needing to be changed after their submittal has been entered into the FENOC Corrective Action Program. As a result of the calculation changes, and a re-examination of the proposed Updated Safety Analysis Report (USAR) markups that were included as Addendum 1 of the RLA, updates of the proposed USAR markups are also provided at the end of this attachment, immediately following the CRDA dose calculation update (31 revised or additional USAR pages (the new pages do not include revised USAR text -they are provided only for the purpose of providing information/context)).
The changes to the LOCA calculation include: 1) Correction of the total elemental iodine removal coefficient in the sprayed region of containment to 20.975 hour0.0113 days <br />0.271 hours <br />0.00161 weeks <br />3.709875e-4 months <br /> 1 , 2) Correction of the equilibrium source term values for Y90, Y91, Y92, Zr95, Zr97, Nb95, and Mo99, 3) Correction of engineered safety feature (ESF) leakage timing so that ESF leakage is assumed throughout the 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of the analysis, 4) Removal of the unnecessary hydrogen mixing system model (sensitivity), 5) Increase of the unfiltered flow rate into the control room during the first 30 minutes to be consistent with the value assumed in the CRDA and main steam line break analyses (10 percent above nominal system intake flow rate), 6) Corrected containment leakage flow rates into the annulus (secondary containment), and 7) Additional changes to improve readability and flow of information.
The change to the CRDA calculation clarifies the description of the iodine releases assumed in the calculation.
COVERSHEET:
Table of Contents OBJECTIVE OR PURPOSE SCOPE OF CALCULATION TABLE OF CONTENTS SUBJECT
SUMMARY
OF RESULTS/CONCLUSIONS LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY IMPACT ON OUTPUT DOCUMENTS DOCUMENT INDEX CALCULATION COMPUTATION (BODY OF CALCULATION):
Page i 1.0 PURPOSE ............
.-....................................................................................................*..................................................
1
2.0 BACKGROUND
...***..*...**.*.*.....*..............*.*..*.*....***.....**..*...*............................*........*.*..**..*..**.**.*..*.**..*...**.....**...*.....***
1 3.0 ACCEPTANCE CRITERIA ***.*...****.*....******.**.****..**.**********..**...*********.***..*****....*.***..******.***********.*********************.******.**********
z 4.0 METHOD OF ANALYSIS *.....*.............*.......*.*...**..*.*...**...**.*........*.*.....*...*..*...*....*..*.********.***********.*****..*......*..*....*****...**
3 5.0 ASSUMPTIONS
...*****.....**.....***............**........*....***....***.......**.....**........*.......*..*...****...****...*.******...***......**...............*...*
3 5.1 CONTROL ROOM EMERGENCY RECIRCULATION SYSTEM (CRERS) *****************************************************************************************************
3 5.2 HYDROGEN MIXING SYSTEM.*************************************************.*******************************************************************************
3 5.3 CONTROL ROOM INLEAKAGE
......................................................................................................................................................
4 5.4 DRYWELLFLOWS
.....................................................................................................................................................................
4 5.5 CONTAINMENT LEAKAGE RATE ...................................................................................................................................................
5 5.6 AEGTS FILTRATION
.................................................................................................................................................................
5 5.7 CONTAINMENT SPRAY ................................................................................................................................... ..........................
5 5.8 ECCS LEAKAGE .......................................................................................................................................................................
6 5.9 MSIV LEAKAGE RATE ***************************************************************************************************************************************************************
6 5.10 BYPASS LEAKAGE .....................................................................................................................................................................
7 5.11 SOURCE TERM RELEASE ************************************************************************************************************************************************************
1 6.0 DESIGN INPUT .......................................................................................
- ................................................................ **** 8 6.1 PLANT GRADE ..................................................................
- ............................................................... ......................................
8 6.2 CORE SOURCE TERMS AND RELEASES ...........................................................................................................................................
8 6.2.J Onset of Gap Release .............................................................................................................................................
JO 6.2.2 Release fractions
....................................................................................................................................................
JO 6.3 SUPPRESSION POOL IODINE RE-EVOLUTION
..................................................................................................................................
10 6.4 DOSE CONVERSION FACTORS ....................................................................................................... *********************************************
10 6.5 ATMOSPHERIC DISPERSION FACTORS ********************************************** , ********************************************************************************************
11 6.6 BREATHING RATE AND OCCUPANCY FACTORS ******************************************************************************************************************************
- . 11 6.7 CONTAINMENTVOlUMES
..........................................................................................................................................................
12 6.8 TECHNICAL SUPPORT CENTER DOSES *******************************************************************************************************************************************
12 6.9 MIXING BElWEEN THE UNSPRAYED CONTAINMENT AND SPRAYED CONTAINMENT
...............................................................................
13 6.10 CONTAINMENT LEAKRATE ..........................................................................................................................................................
13 6.11 LEAKAGE AFTER 24*HouRS ........................................................................................................................................................
14 6.12 MSIV FLOWS *************************************************************************************************************************************************************************
15 6.13 RADIONUCLIDE REMOVAL MECHANISMS