ML20137G202
| ML20137G202 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/27/1997 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20007F796 | List: |
| References | |
| W3F1-97-0061, W3F1-97-61, NUDOCS 9704010350 | |
| Download: ML20137G202 (21) | |
Text
s Enttrgy Operrtions, Inc.
Killona LA 70066-0751 Tel 504 739 6660 Charles M. Dugger c Pres duti. Oteatms W3F1-97-0061 A4.05 PR March 27,1997 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-193 Gentlemen: ~
Pursuant to 10CFR50.90, Entergy Operations Inc. (EOI) hereby proposes to amend Operating License NPF-38 by incorporating the proposed changes identified in the enclosure into the Technical Specifications of Waterford 3 to increase the Spent Fuel .
Pool storage capacity and increase the maximum fuel enrichment from 4.9 w/o (nominal weight percent) to 5.0 w/o U-235. As indicated in Tabb 3.3.1 of Attachment 111 of the Enclosure, BORAL, the neutron absorber for the new racks, has been licensed by the NRC for use in numerous nuclear power plant spent fuel storage applications.
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It is requested that the issuance date for this amendment be no later than January 8, 1998. This approval date is necessary to support a rerack completion date of July, 1998, when Cycle 10 fuel will be arriving on site, and to support the expeditious removal of the existing Boraflex racks.
Please note that Attachment lll of the Enclosure contains >some information that is considered proprietary pursuant to 10CFR2.790. In this regard, EOl requests that Attachment 111 be withheld from public viewing. Attachment IV of the Enclosure is s identical, except that the proprietary information has been removed and replaced by a [fCl /
note of explanation at each location where information has been omitted. EOl offers this additional version for the purposes of public review.
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- Technical Specification Change Request NPF-38-193 i l' W3F1-97-0061 I Page 2 March 27,1997 t
The proposed changes have been evaluated using criteria in 10CFR50.92(c), and it has been determined that the changes involve no significant hazards considerations. The ;
bases for these determinations are described in the attached Enclosure. l If you should have any questions on the above or attached, please contact !
Tim Gaudet at (504) 739-6666 or Roy Prados at (504) 739-6632.
Very truly yours, ;
,r
[ 0.M. Dugger I
( i i
't Vice President, Operations i Waterford 3 l l
CMD/RWP/ssf l
Enclosures:
- NPF-38-193 cc: E.W. Merschoff, NRC Region IV j C.P. Patel, NRC-NRR l R.B. McGehee l N.S. Reynolds NRC Resident inspectors Office j Administrator Radiation Protection Division !
(State of Louisiana) ]'
American Nuclear Insurers 1 .
e
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION in the matter of ) ,
) ,
Entergy Operations, Incorporated ) Docket No. 50-382 :
Waterford 3 Steam Electric Station )
l AFFIDAVIT t
Charles Marshall Dugger, being duly sworn, hereby deposes and says that he is .
Vice President Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change ReqJest NPF-38-193; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
4 N
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Charles Marshall Dugger #\
Vice President Operations - Waterford 3 STATE OF LOUISlANA )
) ss PARISH OF ST. CHARLES ) !
Subscribed and sworn to before me, a Notary Public in and for the Parish and State l above named this 27 " day of Om W1 c H .1997. l l
6 .
g Notary Public My Commission expires " * "
- )
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DESCRIPTION OF PROPOSED CHANGE NPF-38-193 I
The following is a request to amend Operating License NPF-38 by incorporating the !
- proposed changes identified in the Enclosure into the Technical Specifications of
- Waterford 3 to increase the Spent Fuel Pool storage capacity and increase the
- maximum fuel enrichment from 4.9 w/o (nominal weight percent) to 5.0 w/o U-235.
- EXISTING SPECIFICATIONS 1
- See Attachment I PROPOSED SPECIFICATIONS i
- See Attachment II BACKGROUND Weterford 3 received its provisional operating license on December 18,1984. Atthat time, the Spent Fuel Pool was authorized to store 1088 fuel assemblies in 16 spent fuel racks. Current projections, based on expected future spent fuel discharges, indicate that loss of full-core-discharge capability will occur at the end of Cycle 10 in 2000. Operation of Waterford 3 beyond loss of full-core-discharge capability is possible for Cycles 11 and 12 to provide an additional three to four years of operation until 2004.
It is well documented that the future of high level nuclear waste, including spent fuel, is uncertain. The Department of Energy has experienced numerous delays and restarts in providing long term safe storage for high level nuclear waste. Accordingly, in EOl's judgment, it is not likely that a facility will be in operation early enough to avoid loss of full-core-discharge capability. Many utilities are reaching their capacity for Spent Fuel Pool (SFP) storage and seeking on-site spent fuel storage solutions.
However, licensing and permitting of additional storage facilities to supplement existing Spent Fuel Pools is costly and sometimes prohibitive.
EOl has evaluated spent fuel storage alternatives that have been licensed by the NRC and which are currently feasible for use at the Waterford 3 site. The evaluation concludes that reracking the Waterford 3 Spent Fuel Pool and adding additional storage racks in the adjacent Cask Storage Pit and Refueling Canal is currently the 1
most cost-effective alternative. Reracking would provide an increase in storage capacity which would maintain the plant's capability to accommodate a full-core-discharge, through the end of Cycle 19 in 2018 and continued operation for two additional cycles until 2022.
DESCRIPTION OF PROPOSED CHANGES Currently Waterford 3 is required to rerack the Spent Fuel Pool to accommodate a full-core-discharge, through the end of Cycle 19 (2018), This proposed modification q will be accomplished by removing the existing racks in the Spent Fuel Pool and l
replacing them with higher density racks. In order to maximize the storage capacity l
of the Spent Fuel Pool, truncation of the sparger lines is also planned. An offset fuel handling tool will be installed to allow access to some of the storage locations that are adjacent to pool walls. Since the Cask Storage Pit is deeper than the Spent Fuel Pool, platforms will be installed beneath the Region 1 racks to allow installation of these racks at the same elevation as the Region 2 racks. Stiffener plates on the travel guides for gate 3A (which encloses the north side of the Cask Storage Pit) will be removed to eliminate interferences with rack installation.
Additional storage racks are proposed for tne Cask Storage Pit and Refueling Canal.
However, the Refueling Canal storage racks would not be installed until after permanent plant shutdown. The new racks will have a closer assembly to assembly spacing to allow for more fuel storage capability. The new racks will be grouped into
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two design styles, designated as Region 1 and Region 2. Both rack styles will ,
contain Boral as the active neutron absorbing poison. Region 1 racks will be located in the Cask Storage Pit and will have the capability to store 255 fuel assemblies in four separate rack modules. Region 2 will have the capability to store 1849 fuel assemblies in sixteen separate rack modules in the SFP and 294 fuel assemblies in five separate modules in the Refurling Canal. This represents an increase of 1310 assemblies over the current licensed capacity of 1088 fuel assemblies.
The p*oposed Region 1 racks will allow fuel storage for enrichments up to 5.0 w/o U-235, without regard to fuel burnup. The Region 2 racks will have minimum burnup requirements for unrestricted fuel storage. Fuel assemblies not meeting the minimum burnup requirements may also be stored in Region 2 utilizing a checkerboard storage pattern. Of course, Region 1 may also be used for storage of fuel assemblies not meeting the minimum burnup requirements established for Region 2. Region 1 has been sized to ensure that an emergency core offload can be accommodated with some allowance for other fuel assemblies that could also I require Region 1 storage. The new racks which comprise Regions 1 and 2 have Boral absorbers which have been sized to fully shadow the active fuel height of all fuel assembly designs stored in the pool.
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- - -. .. -. .- .- - - - - - __ _ - - - -.. ~.. - ..
i A minimum boron concentration of 1720 parts-per-million (ppm) is maintained in the !
- Spent Fuel Pool whenever a fuel assembly is moved. Soluble boron in the Spent Fuel Pool water will ensure that the reactivity is maintained substantially less than the design limitations. Calculations show that for the most severe accident condition, a soluble boron concentration of 700 ppm boron, in addition to the Boral contained in i
Regions 1 and 2, would be adequate to maintain the K eff less than 0.95.
- To accommodate the proposed increase in capacity and change in maximum fuel l
- enrichment, the Waterford 3 Technical Specifications are required to be modified. ;
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The planned expansion (reracking) will take place during Cycle 9 and will preserve the full-core-discharge capability through the end of Cycle 19. The following
- changes are being proposed for Sections 3/4.9 and 5.0, " Refueling Operations" and ,
" Design Features" respectively, of the Technical Specification:
Revise Technical Specification Section 3.9.7(b) for crane travel in the Fuel
, Handling Building from " Loads in excess of 2000 pounds shall be prohibited ;
from travel over fuel assemblies in the spent fuel pool" to " Loads in excess of 4
2000 pounds shall be prohibited from travel over irradiated fuel assemblies in !
the Fuel Handling Building."
Revise Technical Specification Section 3.9.7 APPLICABILITY from *During movement of fuel assemblies in the fuel handling building, or with fuel 1
assemblies in the spent fuel pool" to "During movement of irradiated fuel i assemblies in the Fuel Handling Building, or with irradiated fuel assemblies in the Fuel Handling Building."
l Revise Technical Specification Section 3.9.7 ACTION (b) from "With loads in d
excess of 2000 pounds over fuel assemblies in the spent fuel pool, place the i crane load in a safe position" to "With loads in excess of 2000 pounds over i j irradiated fuel assemblies in the Fuel Handling Building, place the crane load in a safe position."
- Revise Technical Specification Section 4.9.7.2 from "The electrical interlock i system which prevents crane main hook travel over fuel assemblies in the spent fuel pool shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation" to "The electrical interlock system which prevents crane main hook travel over irradiated fuel assemblies in the Fml Handling Building shall be demonstrated
, OPERABLE within 7 days prior to crane use and at least once per 7 Jays thereafter during crane operation."
, Revise Technical Specification Section 4.9.7.3 from " Administrative controls
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which prevent crane auxiliary hook travel with loads in excess of 2000 pounds over the fuel assemblies in the spent fuel pool shall be enforced during crane operations" to " Administrative controls which prevent crane auxiliary hook f
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travel with loads in excess of 2000 pounds over the irradiated fuel assemblies in the Fuel Handling Building shall be enforced during crane operations."
Revise note on bottom of page 3/4 9-7 from "Not required for movement of new fuel assemblies outside the spent fuel pool" to "Not required for movement of new fuel assemblies outside the spent fuel pool and cask storage pit." ,
- Revise Technical Specification Section 3/4.9.7 from "The restriction on 1 movement of loads in excess of the nominal weight of a fuel assembly, CEA, and associated handling tool over other fuel assemblies in the spent fuel pool ,
ensures that in the event this load is dropped (1) the activity release will be l limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. TI<a assumption is consistent with the activity release assumed in the safety analyses" to "The restriction on movement of londs in excess of the nominal weight of a fuel assembly, CEA, and associated handling tool over irradiated fuel assemblies in the Fuel Handling Building ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a j single fuel assembly, and (2) any possible distortion of fuelin the storage racks will not result in a critical array. This assumption is consistent with the !
activity release assumed in the safety analyses." l Revise Technical Specification Section 5.3.1 to change the maximum i enrichment from "4.9 weight percent U-235" to "5.0 weight percent U-235."
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- Revise Technical Specification Section 5.6.1(b) to change the nominal storage i pitch from "A nominal 10.38 inch center to-center distance between fuel assemblies placed in the spent fuel storage racks" to "A nominal 10.185 inch center to-center distance between fuel assemblies placed in the Region 1 (cask storage pit) spent fuel storage racks."
i Add new Technical Specification Section 5.6.1(c) to state "A nominal 8.692 inch center to-center distance between fuel assemblies placed in the Region 2 (spent fuel pool and refueling canal) spent fuel storage racks, except for the four southern-most racks in the spent fuel pool which have an increased center-to-center nominal distance of 8.892 inches in the north-south direction."
Add new Technical Specification 5.6.1(d) to state "New or partially spent fuel assemblies may be allowed unrestricted storage in Region 1 racks."
Add new Technical Specification 5.6.1(e) to state "New fuel assemblies may be stored in the Region 2 racks provided that they are stored in a
' checkerboard pattern' as illustrated in Figure 5.6-1."
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Add new Technical Specification 5.6.1(f) to state " Partially spent fuel assemblies with a discharge burnup in the ' acceptable range' of Figure 5.6-2 may be allowed unrestricted storage in the Region 2 racks."
Add new Technical Specification 5.6.1(g) to state " Partially spent fuel assemblies with a discharge burnup in the ' unacceptable range' of Figure 5.6-2 may be stored in the Region 2 racks provided that they are stored in a
' checkerboard pattern', as illustrated in Figure 5.6-1, with spent fuel in the
' acceptable range' of Figure 5.6-3." 1 Revise Technical Specification Section 5.6.3 to add the statement "When fuel is being stored in the cask storage pit and/or the refueling canal these areas will also be maintained at +40.0 MSL."
Revise Tech;xal Specification Section 5.6.4 to change the description of the storage capacity. "The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1088 fuel assemblies" will be changed to "The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1849 fuel assemblies in the main pool,255 fuel assemblies in the cask storage pit and after permanent plant shutdown 294 fuel assemblies in the refueling canal."
Attachment I provides a copy of the existing Technical Specifications. The revised i pages are provided in Attachment 11.
SAFETY ASSESSMENT The planned expansion of the storage capacity involves replacing the 16 existing rack modules in the Spent Fuel Pool with 16 new high-density Region 2 modules with a total of 1849 storage cells, installing four new high-density Region 1 modules ,
in the Cask Storage Pit with a total of 255 storage cells, and installing five new high- !
density Region 2 modules in the Refueling Canal with a total of 294 storage cells.
After the expansion, the pool will contain two distirict administratively controlled storage regions. Each region is characterized by a nominal center-to-center spacing of the cells. The new cells will contain a fixed neutron absorber for primary reactivity I control. The new racks will be grouped in Regions 1 and 2. Region 1 can store fuel assemblies up to a nominal 5.0 w/o U-235 enrichment, without restriction on burn-up, !
and is designed to accommodate an emergency full-core-discharge. The Region 2 storage racks will have minimum fuel assembly burn-up and enrichment :
requirements.
The expansion will initially increase the total storage space from 1088 to 2104 fuel :
assemblies. This capacity in the Spent Fuel Pool and Cask Storage Pit would allow operation without loss of full-core-discharge capability until the end of Cycle 19, since l Waterford 3 is projecting a total spent fuel inventory of 1920 assemblies after Refuel l
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- 19. Operation beyond full-core-discharge capability would be possible for two additional cycles. However, operation until the end of plant license (Cycle 22) would r require rack storage for 2268 assemblies if the last two cycles were performed without full-core-discharge capability. An additional 294 storage locations would be i provided if the Refueling Canal rack option is pursued after permanent plant !
shutdown. This would increase the total storage capacity to 2398 fuel assemblies ;
enabling fuel shuffling and Cask Storage Pit rack removal. Removal of these racks would allow placement of a spent fuel cask to facilitate assembly relocation from the pool.
Rack modules in both regions will be free-standing and self supporting. The new modules will be separated by a gap of approximately 1-3/4 inches from one another in Region 1, and a gap of approximately 3/4 inch in Region 2. Along the pool walls, a nominal gap will also be provided which varies between 1-1/8 inch and 3 inches.
With the expanded capacity, the Spent Fuel Pool cooling system will be required to remove an increased heat load while maintaining the pool water temperature below the design limit. The maximum heat load typically develops from the residual heat in the pool after the last full-core-discharge at the end of plant li'e.
The Spent Fuel Pool thermal performance, criticality, and seismic response have -
been re-analyzed considering the increased storage capacity and fuel enrichment. i The results of these analyses have shown that the pool storage systems remain adequate.
The Significant Hazards Consideration (SHC), contained herein, and the attached Licensing Report (Attachment Ill) address the safety issues arising from the proposed modification and revisions to the Technical Specifications . The scope of the technical analysis supporting this evaluation focused mainly on the final configuration of the expanded Spent Fuel Pool storage space, The transition to the final configuration involving some intermediate stages during the pool construction is also included in the evaluation.
Mechanical Design Evaluation The new fuel rack designs have been evaluated with respect to the mechanical and l material qualifications, neutron poison and poison surveillance requirements, fuel l handling qualifications, fuel interfaces, and accident considerations.
i The principal construction materials for the new racks will be SA240 Type 304L stainless steel, or plate stock, and SA564-630 precipitation hardened stainless steel l for the adjustable support spindies. The rack designs, material selection and fabrication process will comply with the applicable ASTM Standards A240, A276, A479 A564 and others, for service in the nuclear and the boric acid environments. ;
The governing quality assurance requirements for fabrication of the racks are 6
compatible with the quality assurance and quality control of 10CFR50, Appendix B requirements.
For primary nuclear criticality control in the new racks, a fixed neutron absorber will ,
be used, integrated within the rack structure. The absorber, trade name Boral, is a !
boron carbide and aluminum-composite sandwich. Boralis chemically inert and has a long history of applications in the Spent Fuel Pool environments where it has maintained its neutron attenuation capability under thermal loads. Boral is manufactured under the control of a quality assursnce program which conforms to the requirements of 10CFR50, Appendix B.
The installation of the new rack modules will preserve space for thermal expansion and seismic movement. The support legs on the racks will allow for remote leveling and alignment of the rack modules to accommodate variations in the floor flatness.
A thick bearing pad will be interposed between the rack pedestals and the floor to distribute the dead load over a wider support area.
The rack structural performance with respect to the impact and tensile loads, as well l
) as the subcritical configuration, has been analyzed. The analysis included an accidental drop of a fuel assembly during movement to a storage location and tensile j loads (vertical and eccentric) on the rack arising from a stuck assembly in the storage cell. It has been shown that these accidents will not invalidate the mechanical design and material selection criteria to safely store spent fuel in a coolable and subcritical configuration in any region The storage rack structural integrity, and thus the fuel configuration, will be maintained. The fuel will retain its structural integrity and remain subcritical. '
Criticality Considerations The new spent fuel racks are designed to maintain the required subcriticality margin when fully loaded with enriched fuel and in unborated water at a temperature corresponding to the highest reactivity. For reactivity controlin both R6gion 1 and 2 racks, Boral panels will be used. They have been sized to fully shadow the active fuel height of all nsembly designs stored in the pool. The panels will be held in place and protected against damage by a stainless steel jacket which will be stitch welded to the cell walls. In Region 1 the panels will be mounted on the outside faces of each cell. In Region 2 the poison will be mounted either on the exterior or on the interior of the cells, in an alternating pattern.
The storage of spent fuelin each region will be controlled by the criteria defining the maximum permissible reactivity. Region 1 will store the most reactive fresh Zircaloy-clad fuel with a maximum nominal enrichment of up to 5.0 w/o U-235, or spent fuel regardless of the burn-up history. These modules have been designed to accommodate an emergency core offload. Region 2 storage will also accommodate fuel of 5.0 w/o enrichment, but will be subject to burnup limits. If the assembly does 7
not meet the requirements for unrestricted storage in Region 2, then it must be stored in a checkerboard pattern.
The NRC guidelines ar.d the ANSI standards specify that the margin of safety for criticality be maintained by having the maximum neutron multiplication factor, Keff less than or equal to 0.95, including uncertainties, for all normal and accident conditions. The analysis has shown that this criterion is always maintained under all postulated accidents. The accidents and malfunctions evaluated included a dropped fuct assembly on top of the fuel rack; impact on criticality of water temperature and density effects; and impact on criticality of eccentric positioning of a fuel assembly within the rack.
Criticality evaluations have also been performed for fuel located in the transfer carriage and containment temporary storage racks. These analyses show that Keff is maintained less than or equal to 0.95, including uncertainties, for all normal and accident conditions
. Thermal Hydraulics and Pool Cooling i
4 A comprehensive thermal-hydraulic evaluation of the expanded Spent Fuel Pool, Cask Storage Pit, and Refueling Canal has been performed to analyze their thermal performance. Evaluations performed for the Spent Fuel Pool cooling systems j conservatively considered a total storage capacity of 2485 assemblies. This capacity
- is beyond the 2398 designed storage locations and is based on projected fuel
- l. discharges through the end of plant license.
The calculation of the bounding long-term decay heat for thermal analysis of the pool was performed in accordance with the provisions of the USNRC Branch Technical Position ASB 9-2; " Residual Decay Energy for Light Water Reactors for Long Term Cooling" and took into account both the past discharges and the predicted future j refueling cycles. The fuel discharge plan considertJ 24-month fuel cycles, including refueling outage durations, to be consistent with past performance and expected operations. Decay heat was conservatively calculated based on spent fuel assemblies experiencing 4.5 Effective Full Power Years (EFPY). The decay heat j was calculated as a two-step process. First, the cumulative decay heat from the old assemblies that will be stored in the pool until the start of the final cycle was calculated. Since this heat load varies very slowly over time, it was assumed to remain constant in the anaiysis. This heat was then added to the time-variant decay heat generated by the final discharge itself, assumed to take place at a high rate of fuel transfer from the vessel to the pool to maximize the heat addition.
The bounding decay heat was used to calculate the time-dependent pool bulk temperature. The local water temperature determinations are performed assuming that the pool is at its peak bulk temperature. The worst location was identified as the cell with the hottest assembly and the most restrictive convective flow. Conservative 8
values for the axial and radial peaking power factors were used. The local analysis '
was ext 6cded to include the effects of a partially blocked exit flow, postulated from an accidentally dropped assembly on top of the rack. In all cases analyzed, the heat transfer model conservatively accounted for an additional resistance from the fouling of the heat transfer surface in the heat exchangers and performance loss due to plugged tubes. The cooling pump performance assumed in the analysis had been previously verified by actual field tests of the cooling system.
The bulk pool temperature analysis determined that the cooling system has sufficient capacity to maintain the temperature below 140*F during normal refueling heat load (1/2 core discharge beginning 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown), and below 155'F during abnormal refueling heat load (full-core-discharge beginning 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown).
Under normal heat load the pool cooling system must remove 33.7X106 BTU /hr.
This represents an increase over the previous heat load of 22.2X106 BTU /hr determined for this condition, as given in FSAR, Table 9.1-3. The normal refueling heat load scenario considers only the least efficient of the two available pool cooling pumps (i.e., single active failure of the most efficient pump).
Under the maximum abnormal heat load the pool cooling system must remove 50.4X106 BTU /hr. This heat is conservatively determined for the end of the plant's final cycle with the decay heat from 2268 previously discharged assemblies and a full-core-discharge of 217 assemblies. In the analysis, the fuel transfer from the vessel starts 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown at a conservatively high rate of four assemblies per hour. The component cooling water is at its maximum permissible temperature of 90*F during the offload, and the normal refueling cooling alignment (i.e., two pumps aligned with the primary heat exchanger) is modeled. The heat rate calculated for the maximum abnormal condition (i.e., full-core-discharge) is lower than the previously calculated value of 55.7X106 BTU /hr. The previously determined heat load is given in Table 9.1-3 of the FSAR and is larger because it assumes that the entire core is discharged instantaneously three days after shutdown.
The calculated maximum local water temperature is determined to be 192*F in the hottest channel and coincides in time with the highest pool bulk temperature. The
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maximum fuel cladding temperature at the same location is calculated to be 230*F. '
These results conservatively assume 100% blockage of the cellinlet hole in the baseplate. The local boiling point at the top of the fuel, based on the minimum water level in the pool as required by the Technical Specifications, is 236*F which indicates that the channel will remain in subcooled flow, thus minimizing the potential for fuel damage.
A loss of-cooling event was also analyzed for all discharge scenarios. The interruption of the cooling to the pool was assumed at the instant when the pool temperature is at the peak. The analysis determined the time when the pool 9
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_ , inventory begins to boil and the resultant maximum evaporation rate from the
} surface. The time to boil is 2.89 hours0.00103 days <br />0.0247 hours <br />1.471561e-4 weeks <br />3.38645e-5 months <br /> after the cooling is lost in the most severe scenario (the ensuing rate of evaporative loss would not result in the fuel being uncovered until after an additional 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />). This period allows sufficient time for the operators to intervene and line up an alternate source of replenishing the pool inventory and removing the decay heat. Restoration procedures have been established for such contingency which direct the operators to restore the power to all vital station load centers as soon as possible. However, should failure to restore power occur before boiling takes place, long term make-up to the pool can be provided from either the Refueling Water Storage Tank, the Condensate Storage Pool or the Fire Water System.
An evaluation of the Fuel Handling Building's heating, ventilation, and air conditioning (HVAC) system under the conditions of maximum pool heat load (normal and abnormal discharge scenarios) was also performed. This evaluation has confirmed the adequacy of the HVAC system.
Seismic and Structural Evaluation
- ; A complete re-evaluation of the mechanical and civil structures, to address the
[ structural issues resulting from the expansion of the pool storage capacity, has been performed. The analysis considered the loads from seismic, thermal, and mechanical forces to determine the margin of safety in the structural integrity of the fuel racks, the Spent Fuel Pool and the liner, and the Fuel Handling Building. The loads, load combinations, and acceptance criteria were based on the ASME Section ill, Subsection NF, and on NUREG-0800, SRP Section 3.8.4, Appendix D.
- a. The storage rack evaluation The final configuration of the pool will consist of free standing and self-supporting Region 1 and Region 2 modules. The seismic analysis is performed using a whole pool multi-rack analysis, it was based on the simulation of the Safe Shutdown Earthquake (SSE) and the Operating Basis Earthquake (OBE)in accordance with SRP 3.7.1 requirements.
Separate models were developed for each of the three distinct storage locations; Spent Fuel Pool, Cask Storage Pit, and Refueling Canal. All rack modules in the Cask Storage Pit and Refueling Canal models were fully loaded with a conservatively high inertial mass of a 1600 i
i pound fuel assembly. The Spent Fuel Pool rack modules were L analyzed as completely full using an increased assembly weight of 2115 pounds. Actual assembly weight is 1517 pounds maximum.
The results indicate that the maximum seismic displacements result in some impacts with the pool walls and some inter-rack impacts. The resultant member and weld stresses in the racks are all below the 10
allowable stresses, with a safety factor of at least 1.14. This safety l factor is for the female pedestal thread shear stress. The minimum safety factor for the cell membrane material and associated welding is ;
1.81. The racks will remain functional during and after a Safe l Shutdown Earthquake. ;
Fatigue analysis was performed on the storage racks to determine the cumulative damage factor resulting from five operating basis earthquakes followed by one design basis earthquake. This analysis showed that the factor of safety is greater than 10 for fatigue within the rack components.
The rack analysis provides pedestal to bearing pad impact loads resulting from lift-off and subsequent resettling during dynamic events.
The pool floor stresses were evaluated for these impact loads and I determined to remain within allowable limits even when considering the pedestal located directly over leak chases.
In addition to the seismic evaluations, the storage racks were also l analyzed for all postulated accident conditions. A fuel handling )
accident involving a fuel assembly dropped from the Fuel Handling l Machine highest possible lift point would not compromise the integrity of the rack. Permanent deformation of the rack would be limited to the top region only. This is acceptable since the rack cross-sectional ,
geometry at the active fuel height is not altered. Thus, the functionality l of the rack is not affected. i l
In the event of a stuck fuel assembly in the rack, the resultant load on the members will not affect the rack structural integrity to maintain the fuel storage qualifications.
- b. Pool and Fuel Handling Building structural evaluation The Fuel Handling Building consists of cast-in-place concrete interior and exterior walls. It is structurally isolated from other structures, except for the basemat which is shared with the Reactor kJxiliary
- Building and the Reactor Containment Building. This common basemat was taken into account during development of the seismic spectra during original plant design and is not affected by the pool expansion.
The Fuel Handling Building is designed as a seismic Class I structure.
The Spent Fuel Pool is a cast-in-place steel lined reinforced concrete tank structure that provides space for storage of spent fuel assemblies.
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The pool structure and appropriate portions of the Fuel Handling Building have been analyzed using a 3-D finite element model seismically accelerated with motion applied at the base mat level and pool rack and hydrodynamic loads applied. The individualloads and load combinations used were in accordance with NUREG-0800, SRP Section 3.8.4 and based on the " ultimate strength" design method. The primary loads considered were:
the dead weight of the concrete structure, fully loaded racks, and the water, seismic motion consistent with the original plant design for 2%
damping of the OBE and SSE cases applied at the base mat levei, hydrostatic pressure force lateral to the walls, hydrodynamic coupling forces applied to the lower portion of the wall and water slosh pressures on top portion of the wall, bounding thermal loads from a full-core-discharge and a loss of
, cooling, producing the largest temperature gradient across the thickness of the wall and the slab, reactive forces on the pool wall from the Fuel Handling Building structure, and seismically induced rack / pool wall impact loads.
In addition to the loads described above, the pool structure and liner were also analyzed for mechanicalloads under accident conditions including drop of a spent fuel cask into the Cask Storage Pit. Analyses were also performed on liner fatigue considering both temperature and t;eismic cycles. The result of the analyses performed on the Spent FLei Pool and Fuel Handling Building indicate that under all postulated loadings the structural components, floor slabs, pool walls, supporting columns, liner and its anchorages will be subjected to stresses or strains within acceptable limits.
Radiological Considerations Radiological consequences of accidents in the Spent Fuel Pool building have been evaluated. The events considered were a fuel handling accident and a rack drop.
The fuel handling accident involves the release of the gaseous fission products contained in the fuel / cladding gaps of 60 fuel rods in a peak-power 16x16 C.E. fuel assembly,72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown. The changes in source term due to the 12
j higher feel enrichment and burnup have been re-evaluated and have been shown to
- be acceptable.
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- A rack drop accident with radiological consequences is unlikely since all rack 4
movement during the removal and installation phase will follow safe load paths that prevent heavy loads from being transported over the stored spent fuel. Thus, there 1
, are no credible radiological consequences from this accident.
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, There has been no steady long-term increase of radiological conditions in the Spent i
Fuel Pool resulting from the radionuclides within the fuel as more spent fuel is added to the pool. The radiological conditions within the building are typically dominated by
, the most recent batch of the spent fuel from a full-core-discharge. The radioactive
- inventory of the older fuel that will increase with the expanded storage capacity will be insignificant compared to that of the recent offload.
l Since the new storage racks will be located in closer proximity to the Spent Fuel Pool walls, an increase in the adjacent radiological doses is expected. Radiological
. analyses have shown that the dose levels adjacent to all pool areas will remain within acceptable levels, r
Supporting Analysis For supplemental information on the Waterford 3 Spent Fuel Pool Reracking f
Proposed License Amendment, refer to the attached Licensing Report. Two
- versions of the report are attached. The version included as Attachment lli contains
- complete documentation for all sections of the report, including some information
} which is considered proprietary pursuant to 10CFR2.790. Entergy Operations Inc.
i (EOl) requests that this version be withheld from public viewing. The version j i included as Attachment IV is identical, except that proprietary information has been removed and replaced by a note of explanation at each location where information has been omitted. EOl offers this additional version for the purposes of public review. )
SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION l j in accordance with 10CFR50.92, EOl has reviewed the proposed changes and has
- concluded that they do not involve a Significant Hazards Consideration (SHC). The i basis for this conclusion is that the three criteria of 10CFR50.92(c) are not l
! compromised. The proposed changes do not involve a SHC because they would j not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
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In the analysis of the safety issues concerning the expanded pool storage capacity, the following potential accident scenarios have been considered:
- a. A spent fuel assembly drop in the Spent Fuel Pool I l
- b. Loss of Spent Fuel Pool cooling flow I 2
- c. A seismic event i
- d. An accidental drop of a fully loaded fuel shipping cask l l
The probability that any of the accidents in the above list can occur is not i significantly increased by the modification itself. All work in the pool area will be ,
controlled and performed in strict accordance with specific written procedures. l Accordingly, the proposed modification does not involve a significant increase in l i the probability of an accident previously evaluated, i The consequences of an accidental drop of a fuel assembly in the Spent Fuel Pool have been evaluated. The results show that such an accident will not 1 distort the racks sufficiently to impair their functionality. The minimum l subcriticality margin, Keff less than or equal to 0.95, will be maintained. The structural damage resulting from a fuel assembly drop remains unchanged. The radiological consequences are not greatly increased due to the changes in fuel enrichment and burnup. Doses will remain below the levels required by j 10CFR100. Thus, the consequences of such an accident remain acceptable
- , and are not different from any previously evaluated accidents that have been reviewed and found acceptable by the NRC.
The consequences of a loss of Spent Fuel Pool cooling have been evaluated and found acceptable. The expansion of the pool storage capacity does not increase the failure modes of the pool cooling system. In the unlikely event that 4 all pool cooling is lost, sufficient time is available for the operators to provide alternate means of cooling before the onset of pool boiling.
The consequences of a design basis seismic event have been evaluated and j found acceptable. The new racks have been analyzed in their new configuration
, and found safe during seismic motion. The structural capability of the pool will not be exceeded under the dead weight, thermal, and seismic loads. The Fuel Handling Building structure will retain the necessary safety margins during a seismic event. Thus, the consequences of a seismic event are not significantly increased.
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I The consequences of a spent fuel cask drop into the Cask Storage Pit have l been analyzed along with the new rack storage configuration. This evaluation i concluded that the floor and walls remain intact with minor local crushing of concrete. The liner plate would sustain limited damage, which is repairable.
Leakage would be limited to flow through the leak chase system and would be collected at the sump. The Fuel Handling Building integrity would not be '
compromised; therefore, there would be no release of contamination outside of the building. Makeup water from the Condensate Storage Pool and/or the Refueling Water Storage Tank would be adequate to offset loss of water inventory due to any leakages.
\
Therefore it is concluded that the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated. !
- 2. Create the possibility of a new or different kind of accident from any previously analyzed.
Due to the new spent fuel storage locations and configurations the following ,
accidents were considered: I
- a. An accidental drop of a rack module during construction activity in the pool
- b. Draining the Cask Storage Pit and Refueling Canal through the floor drains
- c. Fuel assembly mispositioning accident in Region 2.
A construction accident resulting in a rack drop is an unlikely event. Operability of the cranes will be checked and verified before the reracking operation. Alllift rigging and the Fuel Handling Building crane / hoist system will be verified to be in compliance with the provisions of CMAA Specification No. 70, ANSI B30.2, B30.9, and B30.11. All heavy lifts will be performed in accordance with established station procedures, which will comply with NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." This will minimize the possibility of a heavy load drop accident. Safe load paths will be established that will prevent heavy loads from being transported over the spent fuel.
The consequences of an accidental drop of a rack module into the pool during placement have been evaluated. The analysis confirmed that very limited damage to the liner, which could be easily repaired, could occur. All movements of heavy loads over the pool will comply with the applicable guidelines.
Therefore, the consequences of a construction accident are not increased from any previously evaluated accidents.
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The Cask Storage Pit and Refueling canal both have floor drains which will be plugged prior to installation of the new storage racks in each of the respective areas. The plugs will preclude any water loss through the drain system.
Therefore, draining the Cask Storage Pit and Refueling Canal through the floor drains is not a postulated event.
Fuel assembly mispositioning in Region 2 is an unlikely event, since locating :
assemblies which do not meet the burnup criteria will be administratively
. controlled. However, calculations have been performed to evaluate the ;
mispositioning of a single assembly showing that a concentration of 700 ppm '
boron would be adequate to maintain K,n less than 0.95. A boron concentration of 1720 ppm is administratively maintained during fuel movement. Therefore, subcriticality is ensured and the consequences of this accident have been shown to be acceptable.
The change does not alter the operating requirements of the plant or of the equipment credited in the mitigation of the design basis accidents. Therefore, the potential for an unanalyzed accident is not created. The postulated failure 1
modes associated with the change do not significantly decrease the ability to ,
cool the spent fuel in the pool. The resulting structural, thermal, and seismic loads are acceptable.
- 3. Involve a significant reduction in the margin of safety.
The function of the Spent Fuel Pool is to store the fuel assemblies in a subcritical and coolable configuration through all environmental and abnormal loadings, such as an earthquake or fuel assembly drop. The new rack design must meet 1 all applicable requirements for safe storage and be functionally compatible with i the Spent Fuel Pool. l 4
~
EOl has addressed the safety issues related to the expanded pool storage l capacity in the following areas:
- 1. Material, mechanical and structural considerations
- 2. Nuclear criticality
- 3. Thermal-hydraulic and pool cooling l The mechanical, material, and structural designs of the new racks have been reviewed in accordance with the applicable provisions of the NRC Guidance entitled," Review and Acceptance of Spent Fuel Storage and Handling Applications" The rack materials used are compatible with the spent fuel assemblies and the Spent Fuel Pool environment. The design of the new racks preserves the proper margin of safety during abnormalloads such as a dropped j l
16 i
j
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- assembly and iensile loads from a stuck assembly. It has been shown that such loads will not invalidate the mechanical design and material selection to safely l store fuel in a coolable and subcritical configuration.
I
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The methodology used in the criticality analysis of the expanded Spent Fuel Pool
- meets the appropriate NRC guidelines and the ANSI standards. The margin of i safety for subcriticality is maintained by having the neutron multiplication factor
- . equal to, or less than,0.95 under all accident conditions, including uncertainties.
- This criterion has been preserved in all analyzed accidents.
The thermal-hydraulic and cooling evaluation of the pool demonstrated that the J. pool can be maintained below the specified thermal limits under the conditions of )
j the maximum heat load and during all credible accident sequences and seismic events. The pool temperature will not exceed 140*F during the worst single
- failure of a cooling pump. The maximum local water temperature in the hot I channel will remain below the boiling point. The fuel will not undergo any significant heat up after an accidental drop of a fuel assembly on top of the rack j blocking the flow path. A loss of cooling to the pool will allow sufficient time for i
the operators to intervene and line up alternate cooling paths and the means of j inventory make-up before the onset of pool boiling.
- Thus, it is concluded that the changes do not involve a significant reduction in the margin of safety.
- The NRC has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (51FR7751, March 6,1986) of j
~
amendments that are considered not likely to involve a SHC. The proposed changes for Waterford 3 are similar to Example (x): an expansion of the storage capacity of 1 Spent Fuel Pool when all of the following are satisfied:
- (1) The storage expansion method consists of either replacing existing racks with a i design that allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits.
The Waterfc. ' 3 Spent Fuel Pool rerack involves replacement of the existing j racks with a de. ign that will allow closer spacing of the stored fuel assemblies.
l (2) The storage expansion method does not involve rod consolidation or double j tiers.
I I
The Waterford 3 reracking does not involve fuel consolidation. The racks will not j be double tiered; no fuel assemblies will be stored above other assemblies.
I (3) The Keff of the pool is maintained less than, or equal to,0.95.
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The design of the new racks integrates a neutron absorber, Boral, within the racks to allow closer storage of spent fuel assemblies while ensuring that Keff l remains less than 0.95 under all conditions. Additionally, the water in the Spent I I
Fuel Pool does contain boron as further assurance that Keff remains less than 0.95. The boron that is contained in the pool is not credited, except in the accident condition.
.(4) No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to justify the expansion.
The rack vendor has successfully participated in the licensing of numerous other racks of a similar design. The construction process and the analytical techniques of the Waterford 3 pool expansion are substantially the same as in the other completed rerack projects in the industry. Thus, no new or unproven technology is used in the Waterford 3 reracking.
Environmental Considerations i
EOl has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed changes do not significantly increase the types and amounts of effluents that may be released offsite nor significantly increase individual or cumulative occupational radiation exposures.
Based on the foregoing, EOl concludes that the proposed changes meet the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for L an environmental impact statement.
Schedule Regarding the proposed schedule for this amendment, it is requested that issuance be no later than January 8,1998. This approval date is necessary for the expeditious removal of the existing Boraflex racks cod also to support the scheduled rerack completion date of July 1998, which is when Cycle 10 fuel will begin arriving on site. l l
Conclusion i l As discussed herein, the proposed changes to the Technical Specifications do not involve a SHC pursitant to 10CFR50.92. Reracking the Waterford 3 Spent Fuel Pool has been determined to be safe. Additionally, EOl has determined that this license l
- amendment meets the criteria delineated in 10CFR51.22 (c) (9) for a categorical l
! exclusion from the requirements for an environmental impact statement. l l
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