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DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.1 Page 14.1-1 of 14.1-1
14.1 INTRODUCTION
On June 13, 2022, the licensee certified to the NRC that the Palisades Nuclear Plant had both permanently ceased operations (final shutdown May 20, 2022) and that all fuel had been removed from the reactor vessel and placed in the spent fuel pool. Since the facility is no longer authorized by the operating license to enter any operational mode, reactor related accidents, abnormal operational transients, and special events are no longer a possibility.
This chapter discusses: (a) a postulated cask drop accident, (b) a postulated fuel handling accident (FHA) associated with fuel movement until the fuel has been transferred to the Independent Spent Fuel Storage Installation (ISFSI),
(c) a waste liquid incident, and (d) a waste gas incident. Bounding conditions, conservatism in equipment design, conformance to high standards of material and construction, the control of loads and strict administrative controls over facility operations all serve to assure the integrity of the fuel while in the SFP and during fuel transfer to the ISFSI.
A summary Disposition of Events for Palisades is given in Table 14.1-2. This table lists each Chapter 14 event, indicates whether the event is analyzed and provides a reference to the bounding event for events not analyzed. The DSAR section containing a summary of the analyzed events is also given in the table. Table 14.1-3 summarizes the offsite and control room radiological dose consequences and limits for the accidents evaluated in Chapter 14. It is shown that the radiological consequences for Palisades are below the limits specified in 10CFR100.11, the Standard Review Plan, General Design Criterion 19 and/or Regulatory Guide 1.183, as applicable.
Accidents involving fuel and the storage system utilized at the ISFSI are discussed in the storage system Final Safety Analysis Report.
New hazards, new initiators or new accidents that may challenge offsite guideline exposures, may be introduced as a result of certain decommissioning activities. These issues will be evaluated when the scope and type of decommissioning activities are finalized.
14.1.1 DELETED 14.1.2 DELETED 14.1.3 DELETED
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-1 of 14.11-10 14.11 POSTULATED CASK DROP ACCIDENTS 14.11.1 EVENT DESCRIPTION In 2003, Facility Change FC-976 modified the main hoist of the Fuel Building Crane to increase the capacity to 110-tons, and to meet single failure criteria in accordance with NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, and NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants. Postulated load drops from the spent fuel pool area cranes are analyzed, unless the crane and lifting devices are designed and specified to be single failure proof in accordance with NUREG 0612 and NUREG-0554.
Since the main hoist has been upgraded to meet single-failure-proof criteria, analyses of postulated load drops from the main hoist are no longer required.
Although the main hoist of the spent fuel crane is designed and operated in accordance with single-failure-proof criteria for cask handling activities, there may be situations in which lifting devices used with the main hook do not meet these requirements or single-failure-proof features of the main hoist may be disabled. In these situations, the crane no longer meets single failure proof requirements, and load drops are postulated. Therefore, this section contains an outline of the methodology and evaluations used to document the consequences of postulated fuel transfer cask drop accidents in the fuel handling area of the Palisades facility.
The postulated drops of the loaded Sierra Nuclear Multi-Assembly Sealed Basket Transfer Cask (MTC) are the bounding evaluated cask drops, and are assumed to be the result of hypothetical failures in the crane system (main hook) or load handling devices. The load drops were evaluated for structural and radiological effects.
The following is a list of the structural and radiological calculations that were used to provide for the safe use of the loaded MTC in the spent fuel pool area:
EA-FC-864-09 "MSB Transfer Cask drop and Impact Limiter design" (Ref. 4)
EA-FC-864-11 "Evaluation of the MTC/MSB for drop on the VCC with the MTC/MSB C.G. located outside the VCC boundary" (Ref. 5)
EA-FC-864-41 "MTC Over Turn Calculation in the Spent Fuel Pool" (Ref. 7)
NAI-1149-026 Palisades Design Basis Cask Drop Accident AST Radiological Analysis, Revision 0 (Ref. 15)
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-2 of 14.11-10 Since the 15-ton auxiliary hoist of the spent fuel pool crane is not single failure proof, postulated load drops from the auxiliary hoist have been evaluated in accordance with NUREG-0612. Heavy Loads handled with the auxiliary hoist are limited to designated safe load paths. The Bechtel Load Drop analysis for a 25-Ton cask (Reference 1) bounds the structural consequences from any drop from the Auxiliary Hoist.
The evaluation of the radiological consequences for a postulated load drop of a loaded Multi-Assembly Sealed Basket Transfer Cask (MTC) from the main hoist, contained within this section, bounds the radiological consequences from postulated load drops from the auxiliary hoist.
14.11.2 STRUCTURAL ANALYSIS Load drop analyses of a 25-ton cask were performed by Bechtel using the methodology described in Bechtel Topical Report BC-TOP-9 (Reference 1).
The analyses considered 25-ton cask load drops in four specific locations in the spent fuel pool area (References 13). The Bechtel analyses show that energy absorbing material is not required to reduce the amount of energy transferred to the floors of the cask loading area in the spent fuel pool and the washdown pit for this load.
Prior to the upgrading of the spent fuel crane to single failure proof, FC-864 (Reference 2) analyzed drops of a loaded MTC due to an Operating Requirements Manual requirement that movement of loads greater than 25 tons over the main part of the spent fuel pool require an evaluation in accordance with Section 5.1 of NUREG-0612. Postulated drops of the loaded MTC/MSB were performed for the washdown pit, the spent fuel pool cask loading area, specific load transfer configurations in the Track Alley and MTC tip over on to the 11 X 7 fuel racks west of the cask loading area (See Fig 14.11-1). The evaluation concluded that impact limiting pads were required in the cask loading area of the spent fuel pool and the cask washdown pit to reduce the amount of energy transferred to the floors. It should be noted that, since the main hoist is now single-failure-proof, impact limiting pads are not required in these areas, provided the lifting devices and interfacing lift points meet the requirements of NUREG-0612, Section 5.1.6. If any of the single failure features of the main hoist are disabled, impact limiting pads are, therefore, required for lifts greater than 25-tons in the cask loading area of the spent fuel pool and the washdown pit.
A listing of the MTC/MSB cask drop results is in Table 14.11-1.
Heavy load lifts that have not been analyzed are acceptable due to compliance with the requirements of NUREG-0612 and Generic Letter 85-11 in that the probability for load drops is extremely small.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-3 of 14.11-10 14.11.2.1 Analysis Method The previously performed structural evaluation of postulated load drops described within this section only apply if the lifting devices and interfacing lift points do not meet the requirements of NUREG-0612, Section 5.1.6, or if any of the single-failure-proof features of the main hoist are disabled. As discussed in Section 14.11.2, impact limiting pads are not required to reduce the amount of energy transferred to the floors of the cask loading area and the washdown pit for a 25-ton load drop.
14.11.2.1.1 Analysis of Cask Drop Scenarios EA-FC-864-09 was written to address each of the following drop scenarios: Drop of the Multi-Assembly Sealed Basket Transfer Cask (MTC) into the spent fuel pool, into the washdown pit, and on top of the Ventilated Concrete Cask (VCC) which is on to the Load Distribution System (LDS). This analysis also evaluates whether modifications to existing auxiliary building structures are necessary to facilitate cask loading. The following areas required evaluation:
A- The fuel pool cask loading area in the northeast corner of the spent fuel pool was evaluated to determine whether modifications were needed to prevent damage to the spent fuel pool liner in the event that the loaded MTC falls from the crane.
B- The washdown pit was evaluated to determine whether modifications were needed to prevent damage due to a postulated drop of the loaded MTC in to the washdown pit.
C- The Track Alley was evaluated to determine whether modifications were needed for a drop of the loaded MTC/MSB on to the VCC in the Track Alley. Also, the MTC door rails were evaluated to determine the damage to them as a result of the postulated drop of the loaded MTC on to the VCC.
14.11.2.1.2 Cask Overturn Due to Seismic Event For a MTC/MSB cask, EA-FC-864-41 was written to address the impact of an earthquake on the MTC/ MSB during the loading of the MSB and whether the cask could tip over and damage fuel in the adjacent fuel racks.
EA-MOD-2003-019-06 evaluated cask overturning due to a seismic event for the Transnuclear 32PT-S125 Dry Storage Canister (DSC) and OS-197 Transfer Cask (TC). The evaluations addressed the necessity of any modifications to prevent a cask from tipping over in the cask loading area and damaging fuel due to an earthquake.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-4 of 14.11-10 14.11.2.2 Bounding Event Input 14.11.2.2.1 Analysis of Cask Drop Scenarios The important input parameters used in drop analysis EA-FC-864-09 are:
A- Height of the MTC/MSB above the spent fuel pool water level at 646'= 48". The temperature of the spent fuel pool during fuel loading d100°F B- Height of the MTC/MSB above the washdown pit floor at 634' =
192" C- Height of the MTC/MSB above the VCC in Track Alley = 70" The important input parameters called out in EA-FC-864-011 are as follows:
- Drop height = 5.83'
- MTC / MSB weight = 93.5 tons 14.11.2.2.2 Cask Overturn Due to Seismic Event For a MTC/MSB, the important input parameters used in the drop analysis EA-FC-864-41 are:
The Horizontal acceleration = 0.30g The Vertical acceleration = 0.14g For a TC/DSC, the important parameters used in the cask overturn analysis EA-MOD-2003-019-06 are:
The Horizontal acceleration = 0.30g The Vertical acceleration = 0.136g
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-5 of 14.11-10 14.11.2.3 Analysis Results 14.11.2.3.1 Analysis of Cask Drop Scenarios The conclusion of EA-FC-864-09 is as follows:
A- The recessed area in the spent fuel pool northeast corner requires an Impact Limiting Pad. This prevents damage to the spent fuel pool structure and liner.
Note: As discussed in Section 14.11.2, since the main hoist is single failure proof, an impact limiting pad is not required provided the lifting devices meet the requirements of NUREG-0612, Section 5.1.6.
B- The washdown pit requires an Impact Limiting Pad. This prevents damage to the washdown pit floor.
Note: As discussed in Section 14.11.2, since the main hoist is single failure proof, an impact limiting pad is not required provided the lifting devices meet the requirements of NUREG-0612, Section 5.1.6.
C- The Load Distribution System will structurally withstand the drop of a loaded Transfer Cask on to the VCC. Local yielding of the MTC door rail as a result of the MTC drop is acceptable. The MTC can still be lifted and placed in the spent fuel pool and the fuel in the MSB unloaded.
The conclusion from EA-FC-864-11, Revision 3, is that the MTC will not drop below the top of the VCC and that the slab at 649'-0" is capable of resisting the impact as long as the VCC is centered r 6" with the center of the Track Alley hatch.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-6 of 14.11-10 14.11.2.3.2 Cask Overturn Due to Seismic Event For MTC/MSB, the conclusion from EA-FC-864-41 is as follows:
The MTC/MSB fully loaded with fuel or empty will not tip over in the spent fuel pool due to a seismic event of horizontal accelerations of 0.30 gs or less and a vertical acceleration of 0.14 gs or less.
For TC/DSC, the conclusion from EA-MOD-2003-019-06 is as follows:
The TC/DSC fully loaded with fuel or empty will not tip over in the spent fuel pool due to a seismic event of horizontal accelerations of 0.30 gs or less and a vertical acceleration of 0.136 gs or less.
A listing of the cask drop and overturning results is in Table 14.11-1.
14.11.3 RADIOLOGICAL CONSEQUENCES 14.11.3.1 Analysis Method 14.11.3.1.1 Radiological Consequences of a Cask Drop in the Spent Fuel Pool EA-TAM-96-04 updated the offsite radiological dose consequences resulting from a cask drop in the spent fuel pool as a result of C-PAL-96-0789 and C-PAL-96-0956 (References 8 and 9). These condition reports document the possibility of previously unanalyzed unfiltered leak paths from the fuel pool area ventilation. The offsite doses for EA-TAM-96-04 (Reference 3) were acceptable, however, when the control room doses were determined for the release rates of EA-TAM-96-04, it was determined that the time at which operators must switch the CR-HVAC to emergency mode was too short to be practically achievable. Hence, procedure changes were made to eliminate isolatable leak paths prior to heavy load moves over non-fueled areas of the main fuel pool zone. These procedure changes were accounted for in EA-CDA-98-01, which updated the offsite radiological release rates resulting from a cask drop in the spent fuel pool. In addition, EA-CDA-98-01 incorporated new offsite short-term X/Q's per Reference 11 in the determination of offsite dose consequences.
NAI-1149-026 (Reference 15) updated the offsite and onsite radiological consequences to incorporate the alternative source term methodology (Reference 16).
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-7 of 14.11-10 License Condition 2.C.(5) prohibits movement of a fuel cask in or over the spent fuel pool when irradiated fuel assemblies decayed less than 90 days are in the spent fuel pool. A cask drop onto 90 day decayed fuel without the FHB Charcoal Filter operating was analyzed in NAI-1149-026. This scenario assumed all radiation is released unfiltered from the FHB.
Note: The radiological consequences of a cask drop in the spent fuel pool do not take credit for an Impact Limiting Pad in the spent fuel pool.
14.11.3.1.2 Impact on MTC/MSB Due to Postulated Drop on the VCC in the Track Alley EA-FC-864-011 analyzes the resultant impact on the MSB from dropping the loaded MTC on to the VCC. The scenario is that the MTC will first impact the VCC and then will topple and strike the adjacent floor at the 649' elevation.
14.11.3.2 Bounding Event Input 14.11.3.2.1 Radiological Consequences of a Cask Drop in the Spent Fuel Pool The important input parameters used in NAI-1149-026 are given in Tables 14.11-2 and 14.11-3.
100% charcoal filter bypass is assumed for the entire duration of the release since the charcoal filters are not operating. Note that fuel in the MSB does not fail.
14.11.3.3 Analysis Results 14.11.3.3.1 Radiological Consequences of a Cask Drop in the Spent Fuel Pool NAI-1149-026 shows that the MTC/MSB drop meets the acceptance criteria. The results of these calculations are listed in Table 14.1-6.
14.11.3.3.2 Drop of the Loaded MTC on to the VCC in the Track Alley EA-FC-864-011 shows that the MTC/MSB will survive the postulated drop onto the VCC and will also withstand the tipping and coming to rest against the concrete slab at the 649' elevation.
The assessment of the MTC shows that the deflection of the shell, due to impact on the slab at 649'-0", is very small and insignificant.
Therefore, the shell stresses are acceptable. No radiation will be released and the MTC/MSB can be placed back into the spent fuel pool and unloaded.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-8 of 14.11-10 14.
11.4 CONCLUSION
S The structural and radiological evaluations of the impact of dropping the MTC / MSB in the auxiliary building have shown that the postulated drops are within design specifications and regulatory requirements.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-9 of 14.11-10 REFERENCES
- 1. BC-TOP-9 Bechtel Topical Report, Design of Structures for Missile Impact, Revision 1, July 1973 (4320/2236), Revision 2, September 1974 (F567/0239).
- 2. Facility Change FC-864, "Dry Storage of Spent Nuclear Fuel," (G250/0266).
- 3. EA-TAM-96-04 Rev. 0, "Offsite Radiological Dose Consequences of a Cask Drop in the Spent Fuel Pool", October 1996, (G781/2304).
- 4. EA-FC-864-09, "MSB Transfer Cask Drop Analysis and Impact Limiter Design, Rev. 7", September 1995, (G707/2215).
- 5. EA-FC-864-11, "Evaluation of the MTC/MSB drop on the VCC with the MTC/MSB C.G. located outside the VCC boundary, Rev 4", September 1994, (G707/2458).
- 6. EA-FC-864-11, "Evaluation of the MSB for droploads for a hypothetical drop on the Load Distribution System (LDS) in the Track Alley. Rev. 2" April, 1993, (G253/0188).
- 7. EA-FC-864-41, "MTC Overturn Calculation in Spent Fuel Pool, Rev 0", April 1993, (G252/0492).
- 8. C-PAL-96-0789 "Potential Airborne Release in VRS and VRS Barrel Storage Areas," July 18, 1996, (4506/1841).
- 9. C-PAL-96-0956, "Failure to Follow Intent of Refueling & Heavy Load Movement Technical Specifications," August 21, 1996, (4166/0081).
- 10. EA-CDA-98-01, Revision 1, "Offsite Radiological Dose Consequences of a Cask Drop in the Spent Fuel Pool," February 7, 2001, (4931/1906).
- 11. Deleted
- 12. Facility Change FC-976, Modify L-3 to Single Failure Proof and Increase Capacity from 100 Tons to 110 Tons.
- 13. Bechtel Report for 25-Ton Cask, Evaluation of Postulated Cask Drop Accidents, August 1974, (9436/0118).
- 14. EA-MOD-2003-019-06, Evaluation of Decontamination Pit and Spent Fuel Pool Areas for Loads from TN Transfer Cask, Rev. 2, May 2004.
- 15. NAI-1149-026, Revision 0, Palisades Design Basis Cask Drop Accident AST Radiological Analysis, June 2005.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.11 Page 14.11-10 of 14.11-10
- 16. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, July 2000.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.19 Page 14.19-1 of 14.19-5 14.19 FUEL HANDLING ACCIDENT 14.19.1 EVENT DESCRIPTION The possibility of an accident with significant consequences during fuel movement is remote due to the many physical limitations imposed upon fuel movement. Administrative restrictions in fuel movement procedures provide additional margin.
Fuel handling hoists and manipulators are designed so that it is not possible to raise fuel bundles within the spent fuel pool above a position which provides the minimum water shield requirements. In addition to these safeguards, direct radiation monitors at fuel handling areas give operating personnel audible and visual warning of high radiation levels. Fuel storage pool integrity is assured by designing the pool and storage racks as Class 1 structures.
In the spent fuel storage area, the design of storage racks and inspection elevator is such that fuel is always in a subcritical geometrical array based on zero boron concentration in the fuel storage pool water. In addition, during fuel movement, fuel pool water will contain a minimum of 1,720 ppm of boron.
This is well above the 850 ppm required to maintain k-effective d0.95 for two 4.95 wt% Uranium enriched fuel assemblies placed in the tilt machine (Reference 4). Adequate cooling of fuel during handling and storage is supplied by natural convection of the surrounding water. An adequate supply of cooled water is assured by the spent fuel pool cooling system. The fuel handling equipment is described in detail in Section 9.11.
The criticality analysis performed to support crediting soluble boron in the Spent Fuel Pool (SFP) determined that a minimum boron concentration of 850 ppm would provide a k-effective of less than or equal to 0.95. In order to ensure that the design-basis k-effective of 0.95 is not exceeded due to potential dilution events, a boron dilution analysis to support this criticality analysis was also performed (Reference 5). As a result, it was established that a boron concentration of greater than or equal to 1720 ppm provides adequate margin for fuel assembly storage and movement within the SFP.
Based on the creditable dilution events evaluated, the 1 1/2 inch fire hose station is the only system with practically an infinite water storage source (Lake Michigan) that could provide the necessary volume of water needed to dilute the SFP to 850 ppm. However, with this fire hose, it would take over 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to dilute the SFP soluble boron concentration from 1720 ppm to 850 ppm. Thus, if an SFP dilution were to occur from this system, reasonable assurance exists that it would be identified and suppressed by an operator before the 0.95 k-effective limit is reached. As an additional measure, a fuel pool high level alarm was added to give an earlier warning of fuel pool increases which could lead to dilution of the soluble boron concentration.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.19 Page 14.19-2 of 14.19-5 The dilution analysis concluded that an unplanned or inadvertent event that would dilute the SFP is not credible for Palisades. Sufficient time is available to detect and suppress the worst dilution event that can occur from the minimum TS boron concentration to the boron concentration required to maintain the 0.95 k-effective design-basis limit.
Inadvertent criticality or overheating during fuel handling is highly improbable.
Similarly, damage to a fuel bundle as a consequence of external forces is also improbable. Operating procedures prohibit the handling of heavy objects such as shipping casks above the fuel storage rack. Inadvertent disengagement of the fuel bundle from the fuel handling machine is prevented by interlocks; consequently, the probability of dropping and damaging a fuel bundle is low.
14.19.2 THERMAL-HYDRAULIC ANALYSIS A thermal-hydraulic analysis is not applicable for this event.
14.19.3 RADIOLOGICAL CONSEQUENCES 14.19.3.1 Analysis Method For the purpose of defining the upper limit of the consequences of a fuel handling accident, it is assumed that the fuel bundle is dropped during handling. Because of interlocks and procedural and administrative controls, such an event is unlikely. However, if the bundle is damaged to the extent that a number of fuel rods fail, the accumulated fission gases and iodines in the fuel rod gap could be released into the surrounding water. Release of fission products which are not in the gap; i.e., in the fuel matrix, is negligible because the low fuel temperature during fuel handling reduces diffusion through the fuel to an insignificant amount (Reference 1).
The fuel bundles are stored within the spent fuel rack which is an eggcrate structure at the bottom of the spent fuel pool. When the fuel bundles are resting in their normal position within the spent fuel rack, the top of the rack extends above the tops of the stored fuel bundles. Because of the configuration and construction of the spent fuel storage racks, a dropped fuel bundle can strike no more than one fuel bundle in the storage racks. Impact can occur only between the ends of the involved fuel bundles (the bottom end fitting of the dropped fuel bundle striking the top end fitting of a stored fuel bundle).
The fuel handling accident was reanalyzed to account for increased radial power peaking factors and higher fuel burnup levels and to be consistent with the methodology described in Regulatory Guide 1.183, Appendix B, "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident," (Reference 2). Increased radial peaking factors affects the fission product inventory of the peak assembly and increased burnup
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.19 Page 14.19-3 of 14.19-5 levels can affect the amount of some fission products that migrate to the fuel-clad gap.
14.19.3.2 Bounding Event Input To bound the consequences of the event (i.e., Control Room Ventilation Filtration and Spent Fuel Pool Ventilation Filtration are no longer credited),
the fuel handling accident was assumed to occur in the spent fuel pool 17 days after shutdown and involve an entire assembly (216 rods). The FHA analysis takes no credit for control room ventilation filtration or pre-filtration of the stack exhaust via the fuel handling area ventilation system. Instead, unfiltered air is assumed to enter through the normal control room ventilation and fuel handling area ventilation system alignments, and unfiltered air is assumed to exhaust through the stack from the SFP ventilation for the entirety of the event. A control room volume of 35,923 ft3 and unfiltered normal makeup flow rate to the control room of 660.0 cfm throughout the event are assumed. A volume of 10,000 ft3 is assumed for the space above the SFP. Note that crediting normal control room and fuel handling area ventilation is more conservative than assuming no ventilation because the normal ventilation provides a driving force for radionuclide entry into the control room or release via the stack.
The source term used is based on facility operation at 102% of 2,650 MWt. A conservative fuel gap activity based on the peak fuel assembly is assumed to be released to the refueling water. Credit for partial retention of iodine in the refueling water was taken. Significant parameters used in the fuel handling accident analysis are given in Table 14.19-1.
Activity released from the SFP is based on the fission product gap release fractions in Table 14.19-2, per RG 1.183. The iodine released from the assembly gap is assumed to be 99.85% elemental and 0.15% organic, in accordance with RG 1.183. There is a minimum of 22.6 feet of water above the top of the fuel rack in the SFP. Therefore, an effective decontamination factor of 183.07 and an overall decontamination factor for elemental iodine of 252 are used in the FHA analysis.
Table 14.19-3 lists the nuclide inventories used in the revised FHA analysis.
These values assume a design basis power level of 2703 MWt and an initial hour 2-day decay period, and are then decayed an additional 15 days prior to the release.
The atmospheric dispersion factors (X/Q) associated with the transport of released radioactivity are shown in Table 14.19-4. Given the release is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in duration per RG 1.183, only the 0-2 hour timestep is applicable.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.19 Page 14.19-4 of 14.19-5 14.19.3.3 Analysis of Results RG 1.183 dose acceptance criteria for the FHA are a Total Effective Dose Equivalent (TEDE) of 6.3 rem at the Exclusion Area Boundary (EAB) and at the outer boundary of the Low Population Zone (LPZ), assuming a 2-hour release duration. As discussed in RG 1.183, the 10 CFR 50.67 limit of 5 rem TEDE in the control room, for the duration of the accident, applies to all accidents.
As shown in Table 14.3-5, for an FHA in the SFP occurring after an irradiated fuel decay time of at least 17 days, the calculated TEDE values to the control room and EAB are less than the limits set forth in 10 CFR 50.67 and Regulatory Guide 1.183, respectively. For an FHA in the SFP where the radioactivity is released in the first two hours, the ratio of the dispersion factors is such that the dose at the LPZ will be much lower than the calculated dose at the EAB. Therefore, the EAB dose shown in Table 14.19-5 bounds the dose at the LPZ.
14.
19.4 CONCLUSION
S The potential offsite doses resulting from a credible fuel handling accident in the spent fuel pool area are less than the guidelines of Regulatory Guide 1.183 (Reference 2), and the doses to control room personnel are less than the limits provided in 10 CFR 50.67.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.19 Page 14.19-5 of 14.19-5 REFERENCES
- 1. 1980 Palisades FSAR, Chapter 14, Section 19.2.
- 2. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, July 2000.
- 3. No. EA-EC89582-01, Revision 0, Post-Permanent Shutdown Fuel Handling Accident and Waste Gas Tank Rupture, dated April 26, 2021
- 4. EA-SFP-99-03, Rev 0, New Fuel Storage Fuel Pool and Fuel Handling Criticality Safety Analysis, EC0013489.
- 5. EA-WJB-00-01, Revision 2, Spent Fuel Pool Dilution Analysis, EC0044749
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.20 Page 14.20-1 of 14.20-3 14.20 LIQUID WASTE INCIDENT 14.20.1 EVENT DESCRIPTION Accidents which might result in release of activity to the environs involve leakage from the liquid waste system components, or the accidental release to the circulating water discharge canal of the contents of one of the radioactive liquid storage tanks.
14.20.2 THERMAL-HYDRAULIC ANALYSIS A thermal-hydraulic analysis is not applicable for this event.
14.20.3 RADIOLOGICAL CONSEQUENCES 14.20.3.1 Analysis Method Activity release from the liquid waste system to the environs can occur only by accidental discharge to the circulating water discharge canal or by a failure of liquid storage tanks T-90 or T-91.
All components (except tanks T-90 and T-91 discussed below) of the liquid radwaste system are located within the containment and auxiliary buildings.
Piping between the containment and the auxiliary building is run within the containment base slab and within the auxiliary building. Any leakage or spillage from any component will be collected by the floor drains and sumps and will drain to the waste receiver tanks in the liquid waste system. The liquid storage tanks are protected against overpressurization and equipped with level instrumentation and alarms. The floor drains and sumps will contain the leakage or spillage within the containment or auxiliary buildings.
Thus, the leakage or spillage will contaminate only the local area to which it spilled.
Liquid storage tanks T-90 and T-91 are not located within the containment or auxiliary buildings. However, both tanks are equipped with level indication and level alarms, and internal overflows back to the auxiliary building. Both tanks have administrative controls that maintain tank activity concentration within the limits described in Section 11.2.3.1, thereby ensuring that 10 CFR 20 dose limits would not be exceeded in the event of a tank failure.
Evaluation of the credibility of an accidental release of radioactive liquid to the circulating water canal is based on the following analysis of the waste discharge operating procedure, monitoring functions and failure consequence.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.20 Page 14.20-2 of 14.20-3 14.20.3.2 Bounding Event Input In general, the procedure for discharging liquid wastes to the circulating water canal is as follows:
- 1. A batch of waste is collected in one of the waste tanks and the tank is isolated.
- 2. This batch is thoroughly mixed by recirculation and a sample is taken for radiochemical analysis.
- 3. If the sample analysis indicates that the quantity of activity to be released is within permissible limits as defined by 10 CFR 20, the quantity of activity is recorded corresponding to the tank volume and its activity concentration. If the release is not within permissible limits, the contents of the waste tank are retained in the tank or recycled through the treatment system.
- 4. If the activity can be released within permissible limits, three valves downstream of the waste radiation monitor must be opened by operator action. The first valve is a manually operated valve that is normally kept locked closed. The next two valves are air to open, fail closed. The first air operated valve in the flow direction is downstream from and controlled by a radiation monitor in the discharge line. This radiation monitor will automatically trip the valve closed on receipt of a high radiation signal. This monitor is described in Section 11.5. The second air operated valve is interlocked with the dilution water pumps and will close if these pumps are not running for dilution of the discharged wastes.
14.20.3.3 Analysis of Results The discharge of liquid waste is governed by automatic interlock and administrative control. In addition, there is a monitor in the circulating water discharge canal to provide a backup to the discharge process monitor by alarming any discharge to Lake Michigan in excess of 10 CFR 20 limits.
These discharge monitors are provided with a check source to permit testing of a monitor and its circuitry from the control room prior to any liquid waste discharge. Failure of a monitor or its circuitry at any time is annunciated in the control room. Any failure in either the monitor or its circuitry will trip closed the valve in the liquid waste discharge line.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.20 Page 14.20-3 of 14.20-3 14.
20.4 CONCLUSION
S It is concluded from the discussion above that the Facility design and administrative controls ensure that radioactive liquid leakage or spillage will be retained within the facility or within 10 CFR 20 dose limits. Also, administrative controls and automatic interlocks, together with the fail safe design of the instrumentation and control devices, provide assurance against any discharge of liquid wastes to the environs in excess of 10 CFR 20 limits.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.21 Page 14.21-1 of 14.21-4 14.21 WASTE GAS INCIDENT 14.21.1 GAS DECAY TANK RUPTURE 14.21.1.1 Event Description Six gas decay tanks contain compressed radioactive gases from the waste gas surge tank which collects gases vented from the Primary Coolant System, the volume control tank and the liquid waste system. These tanks contain the radioactive gases for decay and subsequent controlled release within the limits of 10 CFR 20. Three of the six decay gas tanks were added during the outage of 1971-1973.
The radioactive gases are compressed to a maximum of 115 psia for storage while allowing all the radioisotopes except Kr-85 and Xe-133 to decay to negligible activity levels.
The addition of the three tanks leaves unchanged the existing analysis for a gas decay tank rupture incident since the source term in the new tanks is the same as it is in the original tanks and administrative procedures require that the tanks be normally isolated from each other. This analysis shows that even under the worst expected meteorological conditions, the offsite doses in the event of a gas decay tank rupture are very low.
The waste gas decay tanks are designed for service at 135 psia and 550°F for expected operation at 115 psia and 90°F. Because components of the Waste Gas System are not subjected to high temperatures or high stresses, a failure or rupture of a gas decay tank is unlikely. However, a rupture of this tank is analyzed to define the radiation dose that would result from a malfunction of the Waste Gas System.
14.21.1.2 Thermal-Hydraulic Analysis A thermal-hydraulic analysis is not applicable for this event.
14.21.1.3 Radiological Consequences 14.21.1.3.1 Analysis Method The waste gas incident is defined as an uncontrolled release of the radioactive contents of the Waste Gas System to the atmosphere. Failure of a gas decay tank or the associated piping will result in the release of this radioactive gas.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.21 Page 14.21-2 of 14.21-4 14.21.1.3.2 Bounding Event Input The activity released from a gas decay tank is assumed to be the maximum amount that would accumulate from operation with 1% failed fuel rods in the Primary Coolant System. This activity is obtained from the noble gases krypton and xenon by assuming no gas release from the gas decay tank between refuelings with an equilibrium core.
The principal active gas source is the bleeding and degassing of effluents from the Primary Coolant System. No carry-over of soluble and particulate fission products is assumed. The contents of the gas decay tank will be only noble gases.
The inventory of one gas decay tank containing all radioactive noble gases resulting from reactor operation with 1% failed fuel (gap inventory) is released to the auxiliary building. This activity is obtained from the noble gases krypton and xenon by assuming no gas release from the gas decay tank between refuelings with an equilibrium core radionuclide inventory. Partial decay is assumed corresponding to the time required, approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, to fill one waste gas decay tank. The release then travels unfiltered to the ventilation stack and then to the atmosphere over a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
This value is conservative as the Facility operational control allows the radioactive gases to be released within limits on a continuous basis.
Some of this activity would normally remain in the primary coolant and there would be more holdup time for decay, and the activity removed from the coolant would be distributed among several tanks.
14.21.1.3.3 Analysis of Results The GDTR accident dose consequences are bounded by those of the revised FHA in the spent fuel pool (Reference 3). Specifically, this is accomplished by comparing two inputs to the dose consequence analysis in the revised FHA analysis to the same inputs in the GDTR analysis. The first is the atmospheric dispersion coefficients; the second is the source term. As both GDTR inputs are bounded by those in the revised FHA analysis, the revised FHA is the bounding scenario (i.e., the FHA dose consequences are more severe) and the GDTR does not require any additional decay time beyond what is calculated for the FHA.
The atmospheric dispersion coefficients for the revised FHA are taken from the Reference 3 calculation. The atmospheric dispersion coefficients for the GDTR are taken from the design basis calculation for control room habitability following accidents with radiological consequences. The atmospheric dispersion coefficients in the control room, site boundary, and the low population zone (LPZ) were compared. All atmospheric dispersion coefficients for the revised FHA bound (i.e., result in more severe dose consequences) the atmospheric dispersion coefficients for a GDTR.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.21 Page 14.21-3 of 14.21-4 A comparison of the source term for the revised FHA and the GDTR is also made. The source term for the revised FHA is taken from the Reference 3 calculation. The source term for the GDTR is taken from the design basis calculation for control room habitability following accidents with radiological consequences. When compared, the FHA source term identified in Reference 3 calculation bounds the GDTR source term on a dose equivalent basis.
14.21.2 DELETED 14.
21.3 CONCLUSION
S As the atmospheric dispersion coefficients and the source term for the revised FHA bound those of the design basis GDTR, the dose consequences calculated for the FHA at an irradiated fuel decay time of 17 days, with no credit for control room ventilation filtration and cooling, bound the dose consequences of a GDTR with the same decay period and no credit for these mitigating systems. With the 17-day irradiated fuel decay time, the GDTR dose consequences remain within requirements of 10 CFR 100.
DSAR CHAPTER 14 - SAFETY ANALYSIS Revision 36 SECTION 14.21 Page 14.21-4 of 14.21-4 REFERENCES
- 1. Deleted
- 2. EA-C-PAL-97-0808A, Revision 0, "Determination of Offsite Atmospheric Dispersion Factors (X/Qs) for Radiological Dose Consequence Analyses,"
(4297/1547).
- 3. Calculation No. EA-EC89582-01, Revision 0, Post-Permanent Shutdown Fuel Handling Accident and Waste Gas Tank Rupture, dated April 26, 2021