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Page 1 of 24 DSAR-14.1 Safety Analysis General Rev 3 Safety Classification: Usage Level:
Safety Information Change No.: EC 69954 Reason for Change: Updated for reduced EAB calculations Preparer: C. Waszak Fort Calhoun Station
DSAR-14.1 Information Use Page 2 of 24 General Rev. 3 Table of Contents
- 14. SAFETY ANALYSIS ..................................................................................................... 4 14.1 General ......................................................................................................................... 4 14.1.1 Identification of Occurrences and Accidents ................................................. 4 14.1.2 Deleted ......................................................................................................... 4 14.1.3 Deleted ......................................................................................................... 4 14.1.4 Radiation Monitoring During Accident Conditions ......................................... 4 14.1.5 Deleted ......................................................................................................... 5 14.1.6 Radiological Consequences Methods for Transients Reanalysis ................. 5 14.1.6.1 General Radiological Analysis Methodology .............................. 6 14.1.6.2 Dose Acceptance Criteria .......................................................... 7 14.1.6.3 Core Inventory FCS Equilibrium Core Inventory (Power Level 1530 MWt) ................................................................................. 8 14.1.6.4 Deleted..................................................................................... 14 14.1.6.5 Gap Fractions for non-LOCA Events ....................................... 14 14.1.6.6 Site Boundary Atmospheric Dispersion Factors ....................... 15 14.1.6.7 Fort Calhoun Control Room Atmospheric Dispersion Factors
/Q (sec/m3) ............................................................................ 17 14.1.6.8 Radiological Consequences Control Room Model ................... 19 14.1.6.9 Dose Calculation Methodology ................................................ 20 14.1.7 Deleted ....................................................................................................... 22 14.1.8 Specific References .................................................................................... 22
DSAR-14.1 Information Use Page 3 of 24 Safety Analysis Rev. 3 List of Tables Table 14.1 Deleted ............................................................................................................ 4 Table 14.1 Deleted ............................................................................................................ 5 Table 14.1 Fort Calhoun Fuel Design Parameter Values for Representative Fuel ............. 5 Table 14.1 Deleted ............................................................................................................. 5 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt) ............................... 9 Table 14.1 Deleted ........................................................................................................... 14 Table 14.1 Deleted ............................................................................................................ 14 Table 14.1 Gap Fractions for non-LOCA Events ............................................................... 14 Table 14.1 Fort Calhoun Site Boundary Atmospheric Dispersion Factors /Q (sec/m3) .. 16 Table 14.1 Fort Calhoun Control Room Atmospheric Dispersion Factors /Q (sec/m3). 19 Table 14.1 Radiological Analysis Assumptions and Key Parameter Values - FCS Control Room ........................................................................................................... 20 Table 14.1 Fort Calhoun Fuel Design Parameter Values for Representative Fuel ............. 5 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt) ............................... 9 Table 14.1 Gap Fractions for non-LOCA Events ............................................................... 14 Table 14.1 Fort Calhoun Site Boundary Atmospheric Dispersion Factors /Q (sec/m3) .. 16 Table 14.1 Fort Calhoun Control Room Atmospheric Dispersion Factors /Q (sec/m3). 19 Table 14.1 Radiological Analysis Assumptions and Key Parameter Values - FCS Control Room .............................................................................................................................. 20
DSAR-14.1 Information Use Page 4 of 24 Safety Analysis Rev. 3
- 14. SAFETY ANALYSIS 14.1 General Earlier sections of this report described and evaluated the reliability of major systems and components of the plant from a safety standpoint. For the Safety Analysis it is assumed that certain incidents may occur notwithstanding the precautions taken to prevent their occurrence. The potential consequences of such occurrences are then examined to determine their effect on the plant, to determine whether the plant design is adequate to minimize the consequences of such occurrences, and to provide assurance that the health and safety of the public is protected from the consequences of even the most severe of the hypothetical accidents analyzed.
14.1.1 Identification of Occurrences and Accidents A number of postulated accidents are considered which do not involve the reactor core or coolant system, but which could involve a release of radioactive material to the environment. They are discussed in Sections 14.18,and 14.20. Analysis of these incidents shows that safeguards incorporated in the plant design would limit any release of radioactive material to inconsequential amounts.
14.1.2 Deleted 14.1.3 Deleted Table 14.1.1 - Deleted 14.1.4 Radiation Monitoring During Accident Conditions Gaseous radioactivity is continuously sampled and monitored from the ventilation discharge duct (RM-062). A swing monitor (RM-052) can also monitor gaseous radioactivity and continuously sample particulates from either the containment building or the ventilation discharge duct. There is no longer any requirement to align RM-052 to containment. A multi- channel area monitoring system is provided to measure radiation levels in the containment and auxiliary building. Additionally, the waste disposal system liquid effluent is continuously monitored. The radiation monitoring equipment, (described in detail in Section 11.2.3) in conjunction with installed process instruments and data from the on-site meteorological tower will be used to monitor, locate, quantify, control and plan releases of radioactivity from the plant during normal operation and following an accident.
14.1.5 Deleted Table 14.1.2 - Deleted
DSAR-14.1 Information Use Page 5 of 24 Safety Analysis Rev. 3 Table 14.1 Fort Calhoun Fuel Design Parameter Values for Representative Fuel Fuel Pellet diameter 0.3805 inch Inner cladding diameter 0.387 inch Outer cladding diameter 0.440 inch Active length 129.3 inch Table 14.1 Deleted 14.1.6 Radiological Consequences Methods Fort Calhoun Station (FCS) voluntarily revised the accident source term used in all of its design basis site boundary and control room dose analyses by implementation of the Alternative Source term (AST) (Reference 14.1-13).
The methodology used in the AST design basis accident analyses also reflects a design basis verification/re-constitution effort that was initiated to support a total upgrade to the AST. Included in this verification process were site boundary (Exclusion Area Boundary-EAB and Low Population Zone-LPZ), and control room atmospheric dispersion factors.
The site boundary and control room dose analyses for the design basis accidents were reanalyzed as part of Reference 14.1-13. A majority of the analyses which were reanalyzed for AST are no longer applicable in the permanently defueled state. The following analyses still applicable to FCS were re-evaluated using the AST (applicable DSAR section identified):
Section 14.18, Fuel Handling Accident (in Spent Fuel Pool ) (FHA)
Section 14.20, Waste Liquid Incident (LWTF)
Note that LWTF was not impacted by implementation of the AST, as there is no accident initiated fuel damage associated with this event. However, it was reanalyzed to maintain consistency in design basis.
FCS has completed a full implementation of the AST as defined in Reference 14.1-14. The worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period dose at the EAB, and the dose at the LPZ for the duration of the release, is calculated for each of the events noted above based on postulated airborne radioactivity releases. This represents the post accident dose to the public due to inhalation and submersion for each of these events. Offsite breathing rates used were as follows: 0-8 hr (3.47E-04 m3/sec), 8-24 hr (1.75E-04 m3/sec),24-720 hr (2.32E-04 m3/sec).
The 0 to 30-day dose to an operator in the control room (CR) due to airborne radioactivity releases is developed for all of the DSAR Chapter 14 design basis accidents. This represents the post accident dose to the operator due to inhalation and submersion.
DSAR-14.1 Information Use Page 6 of 24 Safety Analysis Rev. 3 14.1.6.1 General Radiological Analysis Methodology Except as noted below, the updated FCS accident analyses follow the guidance provided in Regulatory Guide (RG) 1.183 (Reference 14.1-14):
The site boundary and control room dose calculations used breathing rates noted in DG 1081 (Draft Guide to RG 1.183). The impact on dose analyses due to usage of these breathing rates instead of RG 1.183 rates is negligible.
To account for fuel conditions outside the bounds of RG 1.183 (Table 3 footnote), conservative estimates of FCS specific fuel gap fractions are utilized for non-LOCA events. For all non-LOCA events analyzed for NRC submittal in Reference 14.1-30 and any future revisions of non-LOCA events, FCS received permission to use the NUREG/CR-5009 gap fractions (Reference 14.1-32).
RG 1.183 does not address the LWTF. The accident scenarios utilized for this analysis reflects other guidance and/or site specific models. This accident is described in DSAR Section 14.20.
Except as noted, assumptions regarding the occurrence and timing of a Loss of Offsite Power (LOOP) are in accordance with RG 1.183 and are selected with the intent of maximizing doses.
For the following reasons, a LOOP is not assumed with FHA and LWTF. Per Information Notice 93-17 the need to evaluate a design basis event assuming a simultaneous/subsequent LOOP is based on the cause/effect relationship between the two events.
The accidents listed above (i.e. FHAs and LWTF) cannot cause a LOOP. Consequently, following the logic sequence discussed in Information Notice 93-17 relative to the LOCA/LOOP, these analyses do not address the potential of a LOOP.
14.1.6.2 Dose Acceptance Criteria The acceptance criteria for the EAB and LPZ dose is based on 10 CFR Part 50 § 50.67, and RG 1.183:
- 1. An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, should not receive a radiation dose in excess of the accident specific total effective dose equivalent (TEDE) value noted in RG 1.183.
DSAR-14.1 Information Use Page 7 of 24 Safety Analysis Rev. 3
- 2. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), should not receive a radiation dose in excess of the accident specific TEDE value noted in RG 1.183.
For LWTF (airborne) the acceptance criteria is 500 mrem based on the fact that it is considered similar to the previously analyzed Waste Gas Decay Tank Rupture (WGDTF). For WGDTF the acceptance criteria was 500 mrem per Branch Technical Position ETSB 11-5.
The acceptance criteria for the Control Room dose is based on 10 CFR Part 50 § 50.67:
Adequate radiation protection is provided to permit occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
14.1.6.3 Core Inventory FCS Equilibrium Core Inventory (Power Level 1530 MWt)
This information is used to develop the radionuclide inventory for the fuel handling accident and therefore remains valid in the permanently defueled state.
The inventory of fission products in the FCS reactor core for radiological consequences calculations is based on maximum full-power operation of the core at a power level equal to the current licensed rated thermal power including a 2% instrument error per RG 1.49 and current licensed values of fuel enrichment and burnup.
DSAR-14.1 Information Use Page 8 of 24 Safety Analysis Rev. 3 The FCS equilibrium core inventory for radiological consequence calculations is determined from ORIGENS calculations (Reference 14.1-19). A conservative "composite" core inventory for the purposes of radiological accident analyses was created using a radionuclide inventory for 3.5%, 4%, and 5% average enriched cores. The highest activity for each isotope for the above three enrichments is chosen to represent the inventory of that isotope in the "composite" core. Using the highest activities for each isotope is quite conservative. The equilibrium core inventory is calculated based on plant operation at 102% and assuming 18-month fuel cycles and a 22-day shutdown refueling.
The equilibrium core at the end of a fuel cycle is assumed to consist of fuel assemblies with three different burnups, i.e., approximately 1/3 of the core with 530-days burnup cycle, 1/3 of the core with 1060-days burnup and 1/3 of the core with 1590 days burnup. Minor variations in fuel irradiation time and duration of refueling outages will have a slight impact on the estimated inventory of long-lived isotopes in the core. However, these inventory changes will have an insignificant impact on the radiological consequences of postulated accidents.
The core inventory for radiological accident calculations developed by ORIGENS, used the above methodology. The FCS equilibrium core inventory is presented in Table 14.1-5.
DSAR-14.1 Information Use Page 9 of 24 Safety Analysis Rev. 3 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt)
Parent Parent Activity Parent Parent Activity Isotope Relationship Isotope (Curies) Isotope Relationship Isotope (Curies)
AG-110 6.84E+06 PU-239 1.80E+04 PARENT: AG-110M 1.71E+05 PARENT: NP-239 8.64E+08 AG-110M 1.71E+05 2ND PARENT: AM-239 1.77E-01 AG-111 2.51E+06 PU-240 2.35E+04 PARENT: AG-111M 2.52E+06 PARENT: NP-240 1.71E+06 GRANDPARENT: PD-111 2.51E+06 2ND PARENT: NP-240M 0.00E+00 AG-112 1.15E+06 PU-241 6.06E+06 PARENT: PD-112 1.14E+06 PARENT: CM-245 1.60E+05 AM-241 7.28E+03 2ND PARENT: CF-249 7.58E-04 PARENT: PU-241 6.06E+06 PU-242 9.98E+01 2ND PARENT: CM-241 8.06E-01 PARENT: AM-242 3.48E+06 AS-76 8.34E+02 GRANDPARENT: AM-242M 4.70E+02 BA-137M 5.01E+06 RB-86 6.93E+04 PARENT: CS-137 5.27E+06 PARENT: RB-86M 5.71E+03 GRANDPARENT: XE-137 7.69E+07 RB-88 3.25E+07 BA-139 7.55E+07 PARENT: KR-88 3.17E+07 PARENT: CS-139 7.39E+07 GRANDPARENT: BR-88 1.76E+07 GRANDPARENT: XE-139 5.55E+07 RB-89 4.26E+07 BA-140 7.56E+07 PARENT: KR-89 3.98E+07 PARENT: CS-140 6.65E+07 GRANDPARENT: BR-89 1.21E+07 GRANDPARENT: XE-140 3.91E+07 RB-90 3.95+E07 BA-142 6.57E+07 PARENT: KR-90 4.28E+07 PARENT: CS-142 2.98E+07 GRANDPARENT: BR-90 6.49E+06 GRANDPARENT: XE-142 5.77E+06 2ND PARENT: RB-90M 1.21E+07 BR-82 1.26E+05 RB-90M 1.21E+07 PARENT: BR-82M 1.08E+05 PARENT: KR-90 4.28E+07 BR-83 5.31E+06 GRANDPARENT: BR-90 6.49E+06 PARENT: SE-83M 2.68E+06 RH-103M 6.52E+07 2ND PARENT: SE-83 2.47E+06 PARENT: RU-103 6.53E+07 BR-85 1.12E+07 GRANDPARENT: TC-103 6.53E+07 PARENT: SE-85 4.66E+06 RH-105 4.19E+07 CD-115 3.41E+05 PARENT: RH-105M 1.28E+07 PARENT: AG-115 2.39E+05 GRANDPARENT: RU-105 4.52E+07 GRANDPARENT: PD-115 3.01E+05 2ND PARENT: RU-105 4.52E+07 2ND PARENT: AG-115M 9.97E+04 RH-105M 1.28E+07 CD-115M 1.58E+04 PARENT: RU-105 4.52E+07 PARENT: AG-115 2.39E+05 GRANDPARENT: TC-105 4.46E+07 GRANDPARENT: PD-115 3.01E+05 RH-106 2.59E+07 CE-141 6.97E+07 PARENT: RU-106 2.36E+07 PARENT: LA-141 6.91E+07 GRANDPARENT: TC-106 3.21E+07 GRANDPARENT: BA-141 6.85E+07 RN-220 1.58E-01 CE-143 6.57E+07 PARENT: RA-224 1.58E-01 PARENT: LA-143 6.52E+07 GRANDPARENT: TH-228 1.57E-01 GRANDPARENT: BA-143 5.70E+07 RU-103 6.53E+07 CE-144 5.38E+07 PARENT: TC-103 6.53E+07 PARENT: LA-144 5.81E+07 GRANDPARENT: MO-103 6.42E+07 GRANDPARENT: BA-144 4.51E+07 RU-106 2.36E+07
DSAR-14.1 Information Use Page 10 of 24 Safety Analysis Rev. 3 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt)
Parent Parent Activity Parent Parent Activity Isotope Relationship Isotope (Curies) Isotope Relationship Isotope (Curies)
CM-242 2.06E+06 PARENT: TC-106 3.21E+07 PARENT: AM-242 3.48E+06 GRANDPARENT: MO-106 2.13E+07 GRANDPARENT: AM-242M 4.70E+02 SB-122 4.03E+04 CM-244 1.99E+05 PARENT: SB-122M 4.03E+03 PARENT: AM-244 6.18E+06 SB-124 3.13E+04 CS-132 1.56E+03 PARENT: SB-124M 6.80E+02 CS-134 7.02E+06 SB-125 3.62E+05 PARENT: CS-134M 1.59E+06 PARENT: SN-125 2.05E+05 CS-134M 1.59E+06 GRANDPARENT: IN-125 3.39E+05 CS-135M 1.64E+06 2ND PARENT: SN-125M 6.24E+05 CS-136 2.19E+06 SB-127 3.58E+06 CS-137 5.27E+06 PARENT: SN-127 1.44E+06 PARENT: XE-137 7.69E+07 GRANDPARENT: IN-127 7.32E+05 GRANDPARENT: I-137 3.97E+07 2ND PARENT: SN-127M 1.95E+06 CS-138 7.90E+07 SB-129 1.32E+07 PARENT: XE-138 7.33E+07 PARENT: SN-129 5.16E+06 GRANDPARENT: I-138 2.00E+07 GRANDPARENT: IN-129 1.47E+06 CS-139 7.39E+07 2ND PARENT: SN-129M 4.93E+06 PARENT: XE-139 5.55E+07 SB-130 4.39E+06 GRANDPARENT: I-139 1.02E+07 SB-130M 1.84E+07 CS-140 6.65E+07 PARENT: SN-130 1.39E+07 PARENT: XE-140 3.91E+07 SB-131 3.24E+07 GRANDPARENT: I-140 2.57E+06 PARENT: SN-131 1.18E+07 EU-154 3.06E+05 GRANDPARENT: IN-131 4.60E+05 EU-155 1.33E+05 SB-132 1.92E+07 PARENT: SM-155 1.55E+06 PARENT: SN-132 9.48E+06 EU-156 9.60E+06 GRANDPARENT: IN-132 1.22E+05 PARENT: SM-156 9.74E+05 SB-132M 1.87E+07 EU-157 1.20E+06 SB-133 2.74E+07 PARENT: SM-157 6.12E+05 PARENT: SN-133 2.59E+06 EU-158 3.58E+05 SE-83 2.47E+06 EU-159 1.81E+05 PARENT: AS-83 3.35E+06 GA-72 6.81E+02 GRANDPARENT: GE-83 5.67E+05 PARENT: ZN-72 6.79E+02 SM-153 1.84E+07 GD-159 2.45E+05 PARENT: PM-153 3.79E+06 PARENT: EU-159 1.81E+05 SN-121 3.34E+05 GE-77 2.89E+04 PARENT: IN-121M 3.11E+05 PARENT: GE-77M 7.77E+04 GRANDPARENT: CD-121 3.05E+05 GRANDPARENT: GA-77 7.58E+04 2ND PARENT: IN-121 2.95E+04 2ND PARENT: GA-77 7.58E+04 SN-123 2.63E+04 H-3 2.33E+04 PARENT: IN-123 2.71E+05 HO-166 3.40E+03 SN-125 2.05E+05 PARENT: DY-166 1.47E+02 PARENT: IN-125 3.39E+05 I-129 1.54E+00 SN-127 1.44E+06 PARENT: TE-129 1.26E+07 PARENT: IN-127 7.32E+05 GRANDPARENT: TE-129M 2.54E+06 SR-89 4.42E+07 2ND PARENT: TE-129M 2.54E+06 PARENT: RB-89 4.26E+07
DSAR-14.1 Information Use Page 11 of 24 Safety Analysis Rev. 3 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt)
Parent Parent Activity Parent Parent Activity Isotope Relationship Isotope (Curies) Isotope Relationship Isotope (Curies)
I-130 9.28E+05 GRANDPARENT: KR-89 3.98E+07 PARENT: I-130M 4.93E+05 SR-90 4.12E+06 I-131 4.08E+07 PARENT: RB-90 3.95E+07 PARENT: TE-131 3.44E+07 GRANDPARENT: KR-90 4.28E+07 GRANDPARENT: TE-131M 8.18E+06 2ND PARENT: RB-90M 1.21E+07 2ND PARENT: TE-131M 8.18E+06 SR-91 5.48E+07 I-132 5.97E+07 PARENT: RB-91 5.14E+07 PARENT: TE-132 5.87E+07 GRANDPARENT: KR-91 2.94E+07 GRANDPARENT: SB-132 1.92E+07 SR-92 5.74E+07 I-133 8.45E+07 PARENT: RB-92 4.51E+07 PARENT: TE-133 4.63E+07 GRANDPARENT: KR-92 1.56E+07 GRANDPARENT: SB-133 2.74E+07 SR-93 6.38E+07 2ND PARENT: TE-133M 3.81E+07 PARENT: RB-93 3.65E+07 I-134 9.44E+07 GRANDPARENT: KR-93 5.22E+06 PARENT: TE-134 7.69E+07 SR-94 6.31E+07 GRANDPARENT: SB-134 5.06E+06 PARENT: RB-94 1.88E+07 2ND PARENT: I-134M 8.26E+06 GRANDPARENT: KR-94 2.37E+06 I-135 8.02E+07 TB-160 4.12E+04 PARENT: TE-135 4.07E+07 TC-99M 6.80E+07 GRANDPARENT: SB-135 2.26E+06 PARENT: MO-99 7.69E+07 I-136 3.76E+07 GRANDPARENT: NB-99 4.51E+07 PARENT: TE-136 1.85E+07 TC-101 6.95E+07 GRANDPARENT: SB-136 3.56E+05 PARENT: MO-101 6.95E+07 IN-115M 3.41E+05 GRANDPARENT: NB-101 6.60E+07 PARENT: CD-115 3.41E+05 TC-104 5.38E+07 KR-83M 5.34E+06 PARENT: MO-104 5.12E+07 PARENT: BR-83 5.31E+06 GRANDPARENT: NB-104 1.96E+07 GRANDPARENT: SE-83M 2.68E+06 TC-105 4.46E+07 KR-85 4.69E+05 PARENT: MO-105 3.76E+07 PARENT: KR-85M 1.13E+07 TE-127 3.53E+06 GRANDPARENT: BR-85 1.12E+07 PARENT: TE-127M 5.87E+05 2ND PARENT: BR-85 1.12E+07 GRANDPARENT: SB-127 3.58E+06 KR-85M 1.13E+07 2ND PARENT: SB-127 3.58E+06 PARENT: BR-85 1.12E+07 TE-127M 5.87E+05 GRANDPARENT: SE-85 4.66E+06 PARENT: SB-127 3.58E+06 KR-87 2.27E+07 GRANDPARENT: SN-127 1.44E+06 PARENT: BR-87 1.80E+07 TE-129 1.26E+07 GRANDPARENT: SE-87 6.65E+06 PARENT: TE-129M 2.54E+06 KR-88 3.17E+07 GRANDPARENT: SB-129 1.32E+07 PARENT: BR-88 1.76E+07 2ND PARENT: SB-129 1.32E+07 GRANDPARENT: SE-88 3.49E+06 TE-129M 2.54E+06 KR-89 3.98E+07 PARENT: SB-129 1.32E+07 PARENT: BR-89 1.21E+07 GRANDPARENT: SN-129 5.16E+06 GRANDPARENT: SE-89 1.22E+06 TE-131 3.44E+07 KR-90 4.28E+07 PARENT: SB-131 3.24E+07 PARENT: BR-90 6.49E+06 GRANDPARENT: SN-131 1.18E+07
DSAR-14.1 Information Use Page 12 of 24 Safety Analysis Rev. 3 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt)
Parent Parent Activity Parent Parent Activity Isotope Relationship Isotope (Curies) Isotope Relationship Isotope (Curies)
LA-140 7.75E+07 2ND PARENT: TE-131M 8.18E+06 PARENT: BA-140 7.56E+07 TE-131M 8.18E+06 GRANDPARENT: CS-140 6.65E+07 PARENT: SB-131 3.24E+07 LA-141 6.91E+07 GRANDPARENT: SN-131 1.18E+07 PARENT: BA-141 6.85E+07 TE-132 5.87E+07 GRANDPARENT: CS-141 5.09E+07 PARENT: SB-132 1.92E+07 LA-142 6.78E+07 GRANDPARENT: SN-132 9.48E+06 PARENT: BA-142 6.57E+07 TE-133 4.63E+07 GRANDPARENT: CS-142 2.98E+07 PARENT: TE-133M 3.81E+07 LA-143 6.52E+07 GRANDPARENT: SB-133 2.74E+07 PARENT: BA-143 5.70E+07 2ND PARENT: SB-133 2.74E+07 GRANDPARENT: CS-143 1.54E+07 TE-133M 3.81E+07 MO-99 7.69E+07 PARENT: SB-133 2.74E+07 PARENT: NB-99M 3.07E+07 GRANDPARENT: SN-133 2.59E+06 GRANDPARENT: ZR-99 6.97E+07 TE-134 7.69E+07 2ND PARENT: NB-99 4.51E+07 PARENT: SB-134 5.06E+06 MO-101 6.95E+07 GRANDPARENT: SN-134 4.42E+05 PARENT: NB-101 6.60E+07 TH-228 1.57E-01 GRANDPARENT: ZR-101 3.96E+07 GRANDPARENT: RA-228 0.00E+00 NB-95 7.28E+07 XE-131M 5.44E+05 PARENT: ZR-95 7.25E+07 PARENT: I-131 4.08E+07 GRANDPARENT: Y-95 7.01E+07 GRANDPARENT: TE-131M 8.18E+06 2ND PARENT: NB-95M 8.29E+05 XE-133 8.46E+07 NB-95M 8.29E+05 PARENT: I-133 8.45E+07 PARENT: ZR-95 7.25E+07 GRANDPARENT: TE-133M 3.81E+07 GRANDPARENT: Y-95 7.01E+07 2ND PARENT: XE-133M 2.65E+06 NB-97 6.79E+07 XE-133M 2.65E+06 PARENT: NB-97M 6.40E+07 PARENT: I-133 8.45E+07 GRANDPARENT: ZR-97 6.74E+07 GRANDPARENT: TE-133M 3.81E+07 2ND PARENT: ZR-97 6.74E+07 XE-135 3.15E+07 NB-97M 6.40E+07 PARENT: I-135 8.02E+07 PARENT: ZR-97 6.74E+07 GRANDPARENT: TE-135 4.07E+07 GRANDPARENT: Y-97 5.52E+07 2ND PARENT: XE-135M 1.76E+07 ND-147 2.78E+07 XE-135M 1.76E+07 PARENT: PR-147 2.76E+07 PARENT: I-135 8.02E+07 GRANDPARENT: CE-147 2.62E+07 GRANDPARENT: TE-135 4.07E+07 NP-239 8.64E+08 XE-137 7.69E+07 PARENT: AM-243 1.25E+03 PARENT: I-137 3.97E+07 GRANDPARENT: PU-243 1.64E+07 GRANDPARENT: TE-137 6.13E+06 PD-109 1.57E+07 XE-138 7.33E+07 PARENT: RH-109 1.31E+07 PARENT: I-138 2.00E+07 GRANDPARENT: RU-109 1.14E+07 GRANDPARENT: TE-138 1.50E+06 2ND PARENT: PD-109M 8.44E+04 Y-90 4.24E+06 PM-147 8.72E+06 PARENT: SR-90 4.12E+06 PARENT: ND-147 2.78E+07 GRANDPARENT: RB-90 3.95E+07 GRANDPARENT: PR-147 2.76E+07 2ND PARENT: Y-90M 2.11E+02
DSAR-14.1 Information Use Page 13 of 24 Safety Analysis Rev. 3 Table 14.1 FCS Equilibrium Core Inventory (Power Level: 1530 MWt)
Parent Parent Activity Parent Parent Activity Isotope Relationship Isotope (Curies) Isotope Relationship Isotope (Curies)
PM-148 6.95E+06 Y-91 5.64E+07 PARENT: PM-148M 1.36E+06 PARENT: SR-91 5.48E+07 PM-148M 1.36E+06 GRANDPARENT: RB-91 5.14E+07 PM-149 2.35E+07 2ND PARENT: Y-91M 3.18E+07 PARENT: ND-149 1.57E+07 Y-91M 3.18E+07 GRANDPARENT: PR-149 1.46E+07 PARENT: SR-91 5.48E+07 PM-151 8.22E+06 GRANDPARENT: RB-91 5.14E+07 PARENT: ND-151 8.14E+06 Y-92 5.78E+07 GRANDPARENT: PR-151 4.87E+06 PARENT: SR-92 5.74E+07 PR-142 2.48E+06 GRANDPARENT: RB-92 4.51E+07 PR-143 6.43E+07 Y-93 4.33E+07 PARENT: CE-143 6.57E+07 PARENT: SR-93 6.38E+07 GRANDPARENT: LA-143 6.52E+07 GRANDPARENT: RB-93 3.65E+07 PR-144 5.40E+07 Y-94 6.79E+07 PARENT: CE-144 5.38E+07 PARENT: SR-94 6.31E+07 GRANDPARENT: LA-144 5.81E+07 GRANDPARENT: RB-94 1.88E+07 2ND PARENT: PR-144M 7.55E+05 Y-95 7.01E+07 PU-238 1.40E+05 PARENT: SR-95 5.66E+07 GRANDPARENT: CM-238 0.00E+00 GRANDPARENT: RB-95 9.04E+06 2ND PARENT: NP-238 1.54E+07 ZR-95 7.25E+07 PARENT: Y-95 7.01E+07 GRANDPARENT: SR-95 5.66E+07 ZR-97 6.74E+07 PARENT: Y-97 5.52E+07 GRANDPARENT: SR-97 2.12E+07
DSAR-14.1 Information Use Page 14 of 24 Safety Analysis Rev. 3 14.1.6.4 Deleted Table 14.1.6 - Deleted Table 14.1.7 - Deleted 14.1.6.5 Gap Fractions for non-LOCA Events The fraction of fission product inventory in the gap to be used for non-LOCA accidents was originally developed in Reference 14.1-13. For non-LOCA events, the gap fractions developed for FCS had a factor of 2 margin applied compared to RG 1.183 gap release fractions. This margin of 2 was applied to address operation at higher power levels of several high burnup rods at FCS (Reference 14.1-24). The gap fractions were reevaluated for the replacement steam generators and replacement pressurizer (Reference 14.1-30). This allowed use of non-LOCA gap fractions obtained from NUREG/CR-5009 (Reference 14.1-31 and Reference 14.1-32)
The FHA for the Control Room and the EAB were updated for Decommissioning and use the gap fractions from NUREG/CR-5009 (Reference 14.1-31). The FHA for the LPZ was not updated and therefore, continue to use FCS specific gap fractions developed in References 14.1-13 and 14.1-24.
The following table provides the gap fractions as identified in RG 1.183, those that include a factor of 2 margin (which are utilized in FHA for the LPZ) and those based on NUREG/CR-5009 ( which are utilized in the FHA for the CR and EAB).
Table 14.1 Gap Fractions for non-LOCA Events FCS Specific Gap NUREG/CR-5009 RG 1.183 Gap Fractions (References Gap Fractions Nuclide Group Fractions 14.1-13 and 14.1-24) (Reference 14.1-31)
I-131 0.08 0.16 0.12 Kr-85 0.10 0.20 0.30 Other Noble Gases 0.05 0.10 0.10 Other Halogens 0.05 0.10 0.10 Alkali Metals 0.12 0.24 0.17
DSAR-14.1 Information Use Page 15 of 24 Safety Analysis Rev. 3 14.1.6.6 Site Boundary Atmospheric Dispersion Factors Normalized atmospheric dispersion ( /Q) values were calculated for the FCS Low Population Zone (LPZ) for post accident gaseous releases from the Auxiliary Building Stack, Radwaste Processing Building Ventilation Discharge Nozzle, and Auxiliary Building Fresh Air Intake (Reference 14.1-23). Normalized atmospheric dispersion ( /Q) values were calculated for the FCS Exclusion Area Boundary (EAB) for post accident gaseous releases from the Auxiliary Building Stack and Radwaste Processing Building Ventilation Discharge Nozzle (Reference 14.1-26).
Reference 14.1-16 methodology was used for calculation of atmospheric dispersion factors. Reference 14.1-13 provides information on how the dispersion factors were developed and calculated. The following assumptions were made for derivation of the atmospheric dispersion factors:
- 1. The plume centerline from each release is transported directly over the receptor (conservative).All releases are treated as point sources (conservative).
- 3. As the radiological releases are from the Spent Fuel Pool and Liquid Waste Tank, no building wake effect is used in the calculation (conservative).
- 4. All releases are treated as ground-level as there are no release conditions that are high enough to escape the aerodynamic effects of the plant buildings (conservative).
- 5. RG 1.111 "plain" terrain recirculation factors are used in the calculation of the annual average /Q values (conservative).
- 6. The EAB relative to the Radwaste Building Exhaust Nozzle is a circle whose radius is based on the shortest distance from the nozzle to the Auxiliary Building Vent Stack EAB (conservative)
The highest EAB and LPZ /Q values from among all 22.5° downwind sectors for each release/receptor combination and accident period are summarized in Table 14.1-9. The 0.5% sector dependent /Q values are presented with parenthesis indicating worst case downwind sector. Reference 14.1-21 indicates how the meteorological data was used.
DSAR-14.1 Information Use Page 16 of 24 Safety Analysis Rev. 3 Table 14.1 Fort Calhoun Site Boundary Atmospheric Dispersion Factors /Q (sec/m3)
Exclusion Area Boundary Averaging Period Release Point 0-2 hr Aux. Bldg. Vent Stack 9.82E-04 (NE)
Radwaste Bldg. 1.77E-03 (NE)
Exhaust Nozzle Low Population Zone Averaging Period Release Point 0-2 hr 0-8 hr 8-24 hr 24-96 hr 96-720 hr All Releases 2.51E-05(NW) 7.29E-06(NW) 4.83E-06(NW) 1.98E-06(NW) 5.49E-07(NW)
DSAR-14.1 Information Use Page 17 of 24 Safety Analysis Rev. 3 14.1.6.7 Fort Calhoun Control Room Atmospheric Dispersion Factors /Q (sec/m3)
The control room (CR) intake /Q values for the releases, which are listed below, are calculated using the methodology documented in Reference 14.1-13. The control room intake
/Q values for the releases were calculated using the latest version of ARCON96 (Reference 14.1-22). Input data consisted of: hourly on-site meteorological data; release characteristics such as release height, stack radius, stack exit velocity, and stack flow rate; the building area affecting the release; and various receptor parameters such as distance and direction from the release to the control room air intake and intake height.
A continuous temporally representative 5-year period of hourly data from Fort Calhouns meteorological tower (i.e., January 1, 1994 through December 31, 1998) (Reference 14.1-21) was used to develop CR Atmospheric Dispersion Factors. Each hour of data, at a minimum, has a validated wind speed and direction at the 10-meter level and a temperature difference between the 60-and 10- meter levels. For details on the explicit calculations, refer to References 14.1-21 and 14.1-22. The specific release-receptor combinations for which /Q values were calculated are as follows:
- 1. Auxiliary Building Stack to Control Room Air Intake
- 2. Auxiliary Building Fresh Air Intake Vent to Control Room Air Intake
- 3. Radwaste Processing Building Ventilation Discharge Nozzle to Control Room Air Intake
DSAR-14.1 Information Use Page 18 of 24 Safety Analysis Rev. 3 The following assumptions were made for determining Control Room Atmospheric Dispersion Factors:
The plume centerline from each release is conservatively transported directly over the control room air intake.
The containment building area having an effect on the dispersion of the applicable releases is that portion above the auxiliary building roof. The Auxiliary Building Stack releases were considered to be effected by the containment building wake effect. All other releases do not consider building wake effect.
The default wind direction range of 90° centered on the direction that transports the gaseous effluents from the release points to either of the intakes is used in the calculation unless a wider range is indicated by the Murphy & Campe S/D ratio.
The default calm wind speed value of 0.5 m/sec is also used in this calculation. Defaults were based on the computer code used (ARCON 96).
The default values for surface roughness length (i.e., 0.10 meter), representative of the topography in the vicinity of FCS, and sector averaging constant (4.0) are used in the calculation.
All releases are conservatively treated as ground level releases as there are no release conditions that merit categorization as an elevated release (i.e., 2.5 times Containment Building height) at FCS.
The /Q values for all Control Room release-receptor combinations are summarized in Table 14.1-10.
DSAR-14.1 Information Use Page 19 of 24 Safety Analysis Rev. 3 Table 14.1 Fort Calhoun Control Room Atmospheric Dispersion Factors /Q (sec/m3)
Averaging Period Release/Receptor Combination 0-2 hr 2-8 hr 8-24 hr 24-96 hr 96-720 hr Aux. Bldg. Stack/CR Air Intake 3.16E-03 2.37E-03 1.16E-03 8.93E-04 7.15E-04 Aux. Bldg. Air Intake/CR Air Intake 4.61E-04 3.12E-03 2.21E-03 9.58E-04 6.88E-04 Radwaste Nozzle/CR Air Intake 1.05E-03 9.04E-04 4.02E-04 2.84E-04 2.27E-04 14.1.6.8 Radiological Consequences Control Room Model For all FCS Radiological Consequence Calculations, the control room (CR) is modeled as a single region (Reference 14.1-13).
This DSAR section addresses the accident configuration and design basis for the CR. All events analyzed for radiological consequences utilized the design basis configuration.
Table 14.1-11 lists the key assumptions/parameters associated with FCS control room model applied to design basis radiological consequences analyses from Reference 14.1-25.
DSAR-14.1 Information Use Page 20 of 24 Safety Analysis Rev. 3 Table 14.1 Radiological Analysis Assumptions and Key Parameter Values - FCS Control Room Control Room Parameters Free Volume 45,100 ft3 Unfiltered Normal Operation Intake 1000 cfm +/- 10%
Unfiltered Inleakage 38 cfm Occupancy Factors 0-24 hr (1.0) 24-96 hr (0.6)96-720 (0.4)
Operator Breathing Rate 0-30 days (3.47E-04 m3/sec) 14.1.6.9 Dose Calculation Methodology Committed Effective Dose Equivalent (CEDE) from inhalation and Deep Dose Equivalent (DDE) from the submersion due to halogens and noble gases were calculated for the offsite and control room locations (Reference 14.1-13). The CEDE was calculated using ICRP-30 dose conversion factors. The committed doses to other organs due to inhalation of halogens and noble gas daughters were also calculated. The decay and daughter build-up during the activity transport among compartments was included.
DSAR-14.1 Information Use Page 21 of 24 Safety Analysis Rev. 3 Committed Effective Dose Equivalent (CEDE) Inhalation Dose-The dose conversion factors by internal organ type were applied to the activity in the air space of the control room, or at the EAB/LPZ. The exposure was adjusted by the appropriate respiration rate and occupancy factors for the CR dose at each integration interval as follows (Reference 14.1-13):
Dh(j)=A(j) x h(j) x C2 x C3 x CB x CO Where:
Dh(j) = Committed Effective Dose Equivalent (rem) from isotope j A(j)= Integrated Activity (Ci-s/m3) h(j)= Isotope j Committed Effective Dose Equivalent (CEDE) dose conversion factor (mrem/pCi) based on Federal Guidance Report No. 11, September 1988 (Reference 14.1-17)
C2= Unit conversion of 1 x 1012 pCi/Ci C3= Unit conversion of 1 x 10-3 rem/mrem CB= Breathing rate (m3/s)
CO= Occupancy factor Deep Dose Equivalent (DDE) from External Exposure B According to Reference 14.1-14 the Effective Dose Equivalent (EDE) may be used in lieu of DDE in determining the contribution of external dose to the TEDE if the whole body is irradiated uniformly. The EDE in the control room is based on a finite cloud model that addresses buildup and attenuation in the air. The dose equation is based on the dose point being at the center of a hemisphere of the same volume as the control room. The dose rate at that point is calculated as the sum of typical differential shell elements at a radius. The equation utilizes, the integrated activity in the control room air space, the photon energy release rates per energy group from activity airborne in the control room, and ANSI/ANS 6.1.1-1991 "neutron and gamma-ray flux-to-dose rate factors".
DSAR-14.1 Information Use Page 22 of 24 Safety Analysis Rev. 3 The Deep Dose Equivalent at the EAB/LPZ locations was conservatively calculated using the semi-infinite cloud model outlined in TID-24190, Reference 14.1-18, where 1 rad is assumed to be 1 rem.
D (x,y,0) rad = 0.25 EBAR (x,y,0)
EBAR = average gamma released per disintegration (MeV/dis)
(x,y,0) = concentration time integral (Ci-sec/m3) 0.25 = [1.11 x 1.6x10-6 x 3.7x1010]/[1293 x 100 x 2]
Where:
1.11 = ratio of electron densities per gm of tissue to per gm of air 1.6x10-6 (erg/MeV) = number of ergs per MeV 3.7x1010 (dis/sec-Ci) = disintegration rate per Curie 1293 (g/m3) = density of air at S.T.P 100 = ergs per gram per rad 2 = factor for converting an infinite to a semi-infinite cloud 14.1.7 Deleted 14.1.8 Specific References 14.1-1 DELETED 14.1-2 DELETED 14.1-3 DELETED 14.1-4 DELETED 14.1-5 DELETED 14.1-6 DELETED 14.1-7 DELETED 14.1-8 DELETED 14.1-9 DELETED 14.1-10 DELETED
DSAR-14.1 Information Use Page 23 of 24 Safety Analysis Rev. 3 14.1-11 DELETED 14.1-12 DELETED 14.1-13 Application for Amendment of Operating License, (Alternate Source Term), LIC-01-0010, February 7, 2001, OPPD to USNRC Document Control Desk 14.1-14 Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 14.1-15 DELETED 14.1-16 Regulatory Guide 1.145, Revision 1, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants 14.1-17 EPA-520/1-88-020, September 1988, Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion 14.1-18 TID-24190, Air Resources Laboratories, Meteorology and Atomic Energy, July 1968 14.1-19 OPPD Calculation FC06800, Revision 1, Bounding Composite Equilibrium Core Inventory with Initial U-235 Enrichments of 3.5 w/o to 5.0 w/o 14.1-20 DELETED 14.1-21 OPPD Calculation FC06806, Revision 0, OPPD Meteorological Data Conversion to ARCON 96 and SWEC Formats.
14.1-22 OPPD Calculation FC06807, Revision 1, Atmospheric Dispersion Coefficients (/Qs) at the Control Room and TSC Air Intakes for Identified Design Basis Accidents Release Points 14.1-23 OPPD Calculation FC06808, Revision 1, Atmospheric Dispersion Coefficients (/Qs) at the EAB and LPZ for Identified Design Basis Accidents Release Points 14.1-24 OPPD Calculation FC06811, Revision 0, Estimate Gap Fractions Based on FCS Specific Fuel Characteristics
DSAR-14.1 Information Use Page 24 of 24 Safety Analysis Rev. 3 14.1-25 OPPD Calculation FC08557, Rev. 1, Fuel Handling Accident In The Spent Fuel Pool Site Boundary And Control Room Dose 14.1-26 OPPD Calculation FC08790, Revision 0, Atmospheric Dispersion Coefficients (/Qs) at the Decommissioning Exclusion Area Boundary (EAB) for Radiological Releases from the Fort Calhoun Station 14.1-27 DELETED 14.1-28 DELETED 14.1-29 DELETED 14.1-30 OPPD Letter, LIC-05-107, October 31, 2005, Application for Amendment of Operating License, Updated Safety Analysis Report Revision for Radiological Consequences Analysis for Replacement NSSS Components 14.1-31 NUREG/CR-5009, PNL-6258, Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors, February 1988 14.1-32 NRC Letter, NRC-06-0146, October 27, 2006, Fort Calhoun Station, Unit No.1 - Issuance of Amendment RE: Changes to the Updated Safety Analysis Report Related to the Radiological Consequences of Events Affected by the Planned 2006 Replacement of the Steam Generators and Pressurizer (TAC NO.
MC8857) 14.1-33 DELETED