ML20032C211
| ML20032C211 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/05/1981 |
| From: | Rothschild M NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Cole R, Mccollom K, Mark Miller Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8111090477 | |
| Download: ML20032C211 (1) | |
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!!arshall E.11111er, Esq., Chairnan Dr. Kenneth A. l'cCollon Administrative Judge Dean, Division of Engineering Atonic Safety and Licensing Board Architecture and Technology U.S. Nuclear Regulatory Cocmission Oklahoma State University Uashington, DC 20555 Stillwater, OK 74074 Dr. Richard Cole Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Comission Washington, DC 20555 In the l'atter of Texas Utilities Generating Company, et al.
(Comanche Peak Steam Electric' Station, Un'its T and 2)
Docket Nos. 50-445 and 50-446
Dear Administrative Judges:
Enclosed for your information is a copy of the application (as amended) subnitted by Texas Utilities Generating Company pursuant to 10 CFR Part 70 for a license to receive, possess, store, inspect and package for transport special nuclear material required for the operation of Comanche Peak Steam Electric Station, Units 1 and 2.
l At this time, the requested license has not been issued.
It is my l
understanding that, subject to making favorable findings on applicable requirements, the NRC Staff will issue the requested license.
I Distribution:
I Document Management Branch Sincerely, l
OELD Formal Files Shapar/Engelhardt Christenbury/Scinto/0lmstead STreby MKarman Farjorie Ulman Rothschild l
MRothschild/STurk Counsel for NRC Staff Chron
Enclosure:
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Safety and Safeguards An U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Sir:
Texas Utilities Generating Company, as agent for Dallas Power &
Light Company, Texas Electric Service Company, Texas Power & Light Company, Texas Municipal Power Agency and Brazos Electric Power Cooperative, Inc., hereby submits an application for a Special Nuclear Material License. Eight (8) copies of the application, as required by Regulatory Guide 3.15, are enclosed for your review and approval.
The license herein applied for will pemit the receipt, possession, storage, inspection, and preparation for transport of special nuclear material required for the operation of the Comanche Peak Steam Elec-tric Station Unit 1, which will be licensed pursuant to the require-ments of 10 CFR Part 50.
Accordingly, a license fee is not required as specified under 10 CFR Part 170. ll(a) (3).
Your prompt consideration of this Special Nuclear Material License application will be appreciated.
Respectfully submitted, A
R J. Cary RJG:kp Enclosure (s) o3
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TEXAS UTILITIES GENERATING COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 APPLICATION FOR SPECIAL NUCLEAR MATERIAL LICENSE This application is filed pursuant to Title 10, Chapter 1, Code of Federal Regulations, Part 70 for authorization to receive, possess, store, inspect, and package for transport unirradiated nuclear fuel assemblies for Unit 1 of the Comanche Peak Steam Electric Station (CPSES). The term of the Special Nuclear Material License requested is for the period beginning January 1,1981 until receipt of the permanent operating license.
The applicants are Texas Utilities Generating Company (TUGC0), Dallas Power & Light Company (DPL), Texas Electric Service Company (TESCO),
Texas Power & Light Company (TPL), Texas Municipal Power Agency (TMPA) and Brazos Electric Power Cooperative, Inc. (BEPC). DPL, TESCO, TPL, TMPA, and BdPC (collectively the "0wners") respectively own 181/3%,
35 5/6%, 35 5/6%, 6 1/5%, 3 4/5% interest in the station as tenants in common. Neither TUGC0 nor the Owners are owned or controlled by an alien, foreign corporation, or foreign government. TUGC0 is the lead Applicant, and as such acts as agent for the Owners for the design, construction, and operation, as well as in licensing matters, but will have no ownership interest.
The location of the offices and principal officers for TUGC0 and the Owners can be found in the Application of TUGC0 and Owners, Docket Nos.
50-445 and 50-446, for Operating Licenses (Class 103) for the Comanche Peak Steam Electric Station Units 1 and 2.
Communications pursuant to this license application should be sent to:
Mr. R. J. Gary Executive Vice President and General Manager Texas Utilities Generating Company 2001 Bryan Toner Dallas, Texas 75201 6e
1.0 GENERAL INFORMATION 1.1 Reactor and Fuel 1.1.1 General The Comanche Peak Steam Electric Station is located in Somervell County, Texas, 65 air miles southwest of the Dallas - Fort Worth Metropolitan Area in North Central Texas. A detailed description of the geographic location is provided in the CPSES Final Safety Analysis Report (FSAR) Section 2.1.1.1, Docket Nos. 50-445 and 50-446. The Unit 1 construction pemit number is CPPR-126 and the Unit 1 Reporting Identification Symbol as assigned by the Nuclear Regulatory Commission is YGL.
1.1.2 Fuel Assemblies The nuclear fuel assemblies consist of slightly enriched uranium dioxide pellets encased in Zircaloy-4 rods. The zircaloy fuel rods have a nominal outside diameter and wall thickness of 0.374 inches and 0.0225 inches, respectively. Each assembly contains 264 fuel rods, 24 Zircalcy-4 control rod guide thimbles, and 1 Zircaloy-4 instrumentation thimble arranged in a 17 x 17 matrix. The 17 x 17 matrix is maintained by 8 inconel grid assemblies located along the length of the fuel assembly.
The assembly top and bottom nozzles are constructed of stainless steel. The assembly is approximately 160 inches in length with a nominal active fuel length of 144 inches. Each assembly is approximately 8.4 inches square.
1.1.3 Assembly Enrichment and Weights The initial core contains nominal assembly enrichments of 1.60 w/o, 2.40 w/o, and 3.1 w/o U-235. The total uranium weight per assembly is nomially 461 Kg of which less than 15 Kg is contained cs U-235. The total assembly design weight, including structural l
l components, is 665 Kg. The fuel assemblies contain no U-233, Pu, depleted uranium, or thorium.
1.1.4 Total Fuel Assemblies and Uranium The total number of fuel assemblies in the initial core is 193. The total weights of U-235 and uranium are approximately 2,100 Kg and 89,060 Kg, respectively. '
1.2 Storage Conditions 1.2.1 Fuel Storage Area Fuel storage and handling operations will be performed in the Fuel Handling Building. One hundred thirty-two (132) fuel assemblies will be stored in the new fuel storage area.
The remaining 61 fuel assemblies for the initial core will be placed in storage in the spent fuel pool storage racks. The new fuel storage area capacity can be increased to 140 assemblies. Should this upgrading occur, the remaining 53 fuel asseablies will be placed in storage in the spent fuel pool storage racks. Detailed elevation and plan views of the Fuel Handling Building are shown in FSAR Figures 1.E-38 through 1.2-40.
1.2.2 Fuel Storage Area Activities Only those activities which involve new fuel receipt, fuel inspection, and fuel handling and storage are normally conducted in or adjacent to the fuel handling and storage areas. No construction activities which could possibly result in damage to the fuel will be allowed in the fuel storage areas during fuel handling, inspection or storage.
1.2.3 Fuel Handling Building Equipment and Systems The Fuel Handling Building structures, components, equipment, systems, and the design criteria used to assure their structural integrity are described in FSAR Sections 9.1 and 3.8.4.
1.2.4 Fire Alarm and Control Systems As presented in FSAR Figure 9.5-40, the Fuel Handling Building combustible loading classification is low based on tables from the National Fire Protection Association (NFPA) and Copper Life Safety Fire Sprinkler System Handbooks. The barriers separating the fire areas are constructed of concrete block or poured, reinforced concrete, or both with approved fire doors, fire dampers, and penetrations of an equivalent rating. Fire protection will be provided by portable extinguishers, hose stations, and a remote manual deluge system. Fire detection is provided by ionization and flame detectors equipped with both local and reinote alarms. A detailed analysis of the fire protection plans is discussed in FSAR Section 9.5.1.3.5..-
1.2.5 Access Control Only authorized personnel will be allowed to enter the fuel handling and storage areas when special nuclear material is present. Controlled access to these areas will be monitored by utilizing intrusion detection methods to detect unauthorized penetrations or activities and, should any suc'i unauthorized intrusions occur, a watchman or offsite response force will be dispatched.
In addition, procedures will be established as part of a physical security plan for dealing with thefts or threats of theft of the special nuclea material.
If and when such events occur, both local law enforcement agencies and the Nuclear Regulatory Comission will be promptly notified.
1.3 Physical Protection, There will be no U-235 (contained in uranium enriched to 207, or more in the U-235 isotope), U-233, or plutonium at the Comanche Peak Steam Electric Station under this license.
Therefore, the requirements of Section 73.1(b) of 10 CFR 73 for strategic special nuclear material and special nuclear material of moderate strategic significance do not apply.
Pursuant to 10 CFR 73.67(f), a detailed physical security plan for special nuclear material of low strategic significance, applicable for the tenn of this Special Nuclear Material License, will be transmitted under a separate cover due to the proprietary information cmtained therein.
1.5 Transfer of Special Nuclear Material The fuel fabricator, Westinghouse Electric Corporation, will be responsible for shipping the special nuclear material as unirradiated fuel assemblies to the applicant.
TUGC0 will not package unirradiated fuel assemblies for delivery to a carrier for transport, except in the event a damaged or unacceptable fuel assembly is to be returned to Westinghouse Electric Corporation. Packaging will be in accordance with the provisions of 00T regulations and 10 CFR Part 71.
1.5 Financial Protection and Indemnity The applicants will apply for nuclear energy liability insurance with the American Nuclear Insurers in the amount of
$1,000,000 to satisfy the financial protection requirements of 10 CFR Part 140.13. The effective period of insurance coverage will be from the time new fuel is received at the Comanche Peak Steam Electric Station until the first fuel assembly is loaded into the reactor. Proof of such financial protection will be furnished prior to issuance of a Special Nuclear Material License pursuant to this application...
2.0 HEALTH AND SAFETY 2.1 Radiation Control 2.1.1 Training and Experience The training and experience of the Comanche Peak Steam Electric Station Chemistry and Health Physics personnel is described in FSAR Section 13.1.3.2.
2.1.2 Procedures and Equipment New fuel will be checked for radioactive contamination by CPSES Chemistry and Health Physics personnel as part of the new fuel inspection procedure. Swipes or smears will be taken of the fuel in order to obtain a representative sample of the surface contamination of the entire assembly and will be counted for alpha and beta / gamma activity to determine the amount of contaminatior: present.
All new fuel that has not been unloaded or unpacked will be handled as contaminated material with all appropriate radiological controls in effect until such contamination checks are performed.
If the amount of contamination is found to exceed allowable limits, the source of the contamination will be determined and appropriate decontamination steps will be initiated as required.
The CPSES Health Physics Program is outlined in FSAR Section 12.5 and describes the procedures and equipment involved in radiological controls.
2.1.3 Detection Calibration and Testing Testing of the detectors used to measure radioactive contamination on new fuel assemblies will consist of daily checks as required on background radiation, detector efficiency, and the updating of a daily trend plot of detector performance.
In addition, the instrumentation will be calibrated as a minimum on a quarterly basis using appropriate calibration sources.
2.2 Nuclear Criticality Safety After receiving the shipping containers at the plant site, only one metal shipping container with fuel assemblies will be opened at any one time. Each fuel assembly will be removed from its shipping container, inspected, and moved to the fuel storage racks if no defects in the fuel assembly are found.
The fuel storage racks in the new fuel storage area and spent fuel pool are designed for a nominal 121 inch and 16 inch, respectively, center-to-center spacing between fuel assembly storage cells. The design of the fuel storage rack assembly is such that it is impossible to insert the new fuel assemblies in other than prescribed locations, thereby preventing any possibility of accidental criticality.
The racks are designed to withstand normal operating loads as well as Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (0BE) seismic loads meeting ANS Safety Class 3 and ASME B&PV Code,Section III, Appendix XVII requirements. The new fuel and spent fuel storage racks are designed to meet the seismic Category I requirements of NRC Regulatory Guide 1.29, Revision 2, February 1976. The fuel storage racks can withstand the impact of a dropped fuel assembly from the maximum lift height of the fuel handling bridge crane without resulting in an unsafe geometric spacing of fuel assemblies.
Handling equipment capable of carrying loads heavier than a fuel assembly are prevented by interlocks or administrative controls, or both, from traveling over the fuel storage areas.
All surfaces that come into contact with fuel assemblies are made of austenitic stainless steel which is resistant to corrosion.
New fuel will be stored in racks in both the new fuel storage area and the spent fuel pool. The racks are designed to prevent accidental criticality even if unborated water is present by maintaining an adequate center-to-center spacing between fuel assemblies. For the flooded condition assuming new fuel of the highest anticipated enrichment (3.5 w/o U-235) in either the new fuel storage racks or the spent fuel pool storage racks, the effective multiplication factor does not exceed 0.95.
Assuming the presence of those possible sources of moderation that could arise during fire fighting operations, the effective multiplication factor does not exceed 0.98 with fuel of the highest anticipated enrichment in the new fuel storage racks. To ensure that the effective multiplication factor for dry storage of new fuel assemblias in the spent fuel pool racks remains less than 0.98, the following administrative procedures will be established:
1.
Dry storage of new fuel assemblies in the spent fuel pool racks will be in a " checker-board" array so that an open storage cell exists on the four adjacent sides of each assembly. Therefore, no two assemblies will be closer than the 21 inch center-to-center spacing of the new fuel storage racks.
2.
The plastic covering around each assembly will be opened at the bottom to allow water drainage should.-
flooding and then draining of the fuel stcrage area occur.
In the criticality analysis of the storage facilities, the fuel assemblies are assumed to be in the most reactive condition with no control rods or removable neutron absorbers present. The assemblies will not be closer together than the design separation provided by the storage racks. Detailed explanations of the criticality safety studies and their results are presented in FSAR Sections 9.1 and 4.3.2.6.
New fuel elements will be removed from their usual storage locations from time to time for such activities as fuel assembly relocation in storage and fuel inspection. The 4
manipulation of the new fuel assemblies wfil be performed by CPSES operations personnel trained in proper fuel handling techniques and, in addition, will use fuel handling procedures which contain provisions to assure that fuel assemblies are handled correctly. Equipment and structures used for fuel handling activities are designed to provide for safe operation as described in FSAR Section 9.1.4.
In order to prevent accidental nuclear criticality, only one new fuel assembly will be allowed to be removed from an approved storage location at any one time. Further discussion of the criticality of fuel assemblies is found in FSAR Section 4.3.2.6.
Because of the fuel storage facilities design and acministrative controls limiting the maximum number of fuel assemblies allowed out of the storage locations, the possibility of accidental criticality during receipt, inspection, and other handling acti<ities is eliminated.
Therefore, an exemption in whole from the requirements of 10 CFR 70.24 is requested as provided by 10 CFR 70.24(d).
2.3 Accident Analysis Interlocks or administrative controls, or both, prevent the Fuel Building handling equipment capable of carrying loads heavier than a fuel assembly from traveling over the fuel storage area. The fuel storage racks are designed to maintain a safe geometric spacing of fuel assemblies despite the impact of a fuel assembly cropped from the maximum lift height of the fuel handling bridge crane.
Since the license will involve only the handling and storage of unirradiated reactor fuel, there would be no significant safety hazard as a result of a fuel handling accident due to the absence of any fission products in the fuel handling areas. Should a fuel handling accident result in the release of any of the uranium contents of the new fuel, Chemistry and Health Physics personnel would be responsible for implementing the appropriate radiological protection measures.
3.0 OTHER MATEPIALS REQUIRING NRC LICENSE There are no special nuclear materials other than that contained in the fuel assemblies requiring an NRC license under this application.
IN WITNESS WHERE0F, TEXAS UTILITIES GENERATING COMPANY has caused this application to be signed in its name by its duly authorized officer this tova day of Juw
, 1980.
TEXAS UTILITIES GENERATING COMPANY By:
h R. (/ Gary
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Executive Vice President and General Manager
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Dear Mr. Ket:
lach:
Your letter dated January 14, 1981, requested additional information in order to complete the review of our applica-tion dated July 10, 1980, for authorization for receipt, possession, storage,' inspection and preparation for trans-port of special nuclear material required for the operation of the Comanche Peak Steam Electric Station Unit 1.
Eight (8) copies of the responses to your comments concerning the required additional information are enclosed for your evaluation.
Your prompt completion of the application review will be appreciated.
Respect fully,
I R. J. Gary RJG:kp Enclosure (s)
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O, RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON TEXAS UTILITIES GENERATING COMPANY LICENSE APPLICATION FOR FUEL STORAGE ONLY AT UNIT 1 COMANCHE PEAK STEAM ELECTRIC STATION Section 1.1.2 Corment (1):
Specify the pellet diameter and fuel rod pitch in an assembly.
Response
The average outside diameter of the slightly enriched uranium dioxide pellets is 0.3225 inches.
The fuel rod pitch in a fuel assembly is 0.496 inches.
Commen t (2):
Confirm 3.1% U235 is the maximum enrichment in a fuel arsembly.
Response
Disregarding the nominal unrichment variations produced during the U. S. Department of Energy's enriching process, the maximum fuel assembly enrichment to be stored under the Comanche Peak Steam Electric Stat.i an Unit 1 Special Nuclear Material License is 3.1 weight percent (w/o)
The highest anticipated enrichment assumed for nuclear criticality safety analyses is 3.5 w/o U-235.
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Section 1.2._2_
Comment (1):
Specify the minimum spacings between fuel handling (e.g., carrier unloading, assembly inspection) and fuel assembly storage ' areas.
Response
The minimum spacing between the nearest fuel storage racks and the new fuel receipt and inspection area is a horizontal distance of approximately 12 feet.
It should be noted, however, that the operating level over the new-fuel storage racks is at elevation 860' whereas the operating level for the new fuel receipt and inspection area is at elevation 841'.
The attached Figure 1 shows the Fuel Building location of the fuel storage areas and the new fuel receipt and inspection area.
Comment (2):
Confirm preoperational testing for_necessary-fuel handling and support systems will be completed prior to receipt of unirradiated fuel. All equipment should be inspected and tested for safe operation prior to use in fuel handling activities.
Response
Prior to receipt of new fuel, all required fuel.
handling equipment and storage.f acilities will be inspected and tested to ensure safe operation during fuel handling activities.
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m Section-1.2.4 Comment (1):
Confirm all construction related to fire protection of the fuel handling and storage areas is completed prior to receipt of unirradiated fuel.
Response
APCSB 9.5-1, Appendix A, requires that'the fire protection program (plans, personnel and equip-ment) for buildings storing new reactor fuel and for adjacent fire zones which could affect the fuel storage zone to be fully operational before fuel is received at the site. Therefore, the Fire Protection Programs for the Fuel Handling Building will be in effect and the suppression systems operational prior to receipt of fuel on the site as stated in the Comanche Peak Steam Electric Station Final Safety Analysis Report (CPSES FSAR) Section 9.5.
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Section 2.1.2 Comment (1):
Identify the responsibility. for administration controls and for the development and approval requirements to ensure safety for all fuel handling and storage operations. Specify the approval requirements.
Response
Administrative controls which. govern the, safe
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handling and storage of fuel will be the respon-sibility.of the Engineering. Superintendent. Those c-procedures which control the safe handling of fuel.
a will be approved by the Station Operations Review
'l Committee (SORC). The function of the SORC'is described in Section 13.4.1 of the CPSES FSAR.
":?j Comment (2):
Confirm all fuel handling operatirns.will be 3.
performed in accordance with approved; written-procedures.
Response
The manipulation of the new fuel assemblies will A
be performed by CPSES Operations personnel trained in proper fuel handling techniques and, in addition, x
will be done in accordance with' approved written fuel handling procedures containing provisions to assure that all fuel assemblies are handled correctly.
Comment (3):
.Specify the precautions taken to meet Al;3RA.
Response
Radiation and contamination monitoring will be performed prior to the initial. handling and storage of new fuel. This practice should identify any possible radiation hazards associated with exteinal contamination of new' fuel assemblies and allow proper.
planning for ALARA controls deemed necessary by Chem-istry and Health Physics personnel. l(Reference.Section 2.1, CPSES Application for Special Nuclear Material License)
Due to the fact that the fuel will be unirradiated, b
there will be no significant radiation hazard' associated with the low level' radioactivity of the fuel itself. The handling and storage of the fuel, as out, lined in the CPSES Application for Special Nuclear Materials License, will be sufficient to maintain radiation exposures,ALARA,.
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s Section 2.2 Comment (1):
Confirm there will be no more than one fuel
~ assembly outside the shipping container or storage rack at any one time.
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Response
In order to prevent accidental criticality, only one nee fuel assembly will be allowed to be re-x moved from a shipping container or an approved
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storage location at any one time.
s Como:ent (2):
Describe the design of the new and spent fuel
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A storage racia. Include sketches in support of the description.
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Response
The,new fuel storage racks (Figure 2) are composed s
of individual. vertical cells fastened together to 1 form'a. module which can be firmly bolted to anchors -
in the ' floor of the new fuel storage pit. The new
" fuel storaZe racks are designed to include" storage for '
two-thirds core at a center-to-center spacing of 21 sinches. The design of the fuel storage rack assembly is such that it is impossible to insert the new fuel assemblies in other than prescribed locations. All surfaces that come into contact with the fuel assemblies are made of annealed austenitic stainless steel. The fuel < storage racks are designed to withstand normal operating loads. as well as Safe Shutdown Earthquake (SSE) and Operatind Basis Earthquake (OBE) seismic loads meeting ANS Safety Class 3 and ASME B&PV Code Section III, Appendix XVII requirements. The fuel storage racks are also designed to meet the seismic p,
Category I requirements of NRC Regulatory Guide 1.29, N
Revision 2, February 1976. The fuel storage racks can withstand an uplift force equal to the 5000 lb. uplift force of the fuel handling bridge crane.
The spent fuel storage racks (Figure 3) are composed of individual vertical cells fastened together at a 16 inch center-to-center spacing to form a module which is firmly bolted to anchors in the floor of the spent fuel pool.
Space between storage cells is bl.ocked ta prevent insertion of fuel. All surfaces that come into contact with fuel assemblies are made of annealed austenitic stainless steel.
The spent fuel storage racks are designed to withstand shipping, handling, normal operating loads (dead loads of fuel assemblies), as well as SSE loads. These racks meet-ANS Safety Class 3 and ASME B&PV Code,Section III, Appendix XVII requirements. The spent fuel storage racks are also
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designed to meet the seismic Category I requirements of Reg.
Guide 1.29, Revision 2, February 1976. The racks can with-stand an uplift force equal to the uplift force of the spent fuel pool bridge hoist.
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Comment (3):
Provide the results of the nuclear criticality safety analyses, the method of analysis used, and the assumptions made in the analysis (e.g.,
degrees of water moderation, dropping of a crane loading on top of the rack).
Response
New fuel is stored in 21 inch center-to-center racks in the new fuel storage facility with no water present.
These racks are designed to prevent accidental critical-ity even if unborated water is present. The design of the new fuel storage racks is such that the effective multiplication factor (keff) does not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry or flooded con-ditions).
For the normally dry condition, kef f does not exceed 0.98 with fuel of the highest anticipated enrich-ment in place assuming possible sources of moderation such as those that could arise during fire fighting operations (such as foam or water mist). Consideration is given to the inherent neutron absorbing effect of the materials of construction.
The new fuel storage racks have adequate energy absorp-tion capabilities to withstand the impact of a dropped fuel assembly from the maximum lift height of the fuel handling bridge crane. An analysis was done using a standard 17 x 17 fuel assembly with the handling tool and a total mass of 2000 lb. falling from a height of 3.5 feet (without damping or energy dissipation) on to the top of a fuel cell. The analysis results show that the fuel cell deforms in compression and shortens in length. The accident would not result in an unsafe geometric spacing of fuel assemblies. Cranes capable of carrying loads heavier than a fuel assembly will be prevented by interlocks or administrative controls, or both, from traveling over the new fuel storage area when new fuel is stored in the new fuel storage racks.
Unborated water of 1.0 gm/cm is assumed in the analysis of new fuel stored in the spent fuel pool racks. Over the range of water densities of interest (corresponding of 60 0
F througt 212 F), full density water is a conservative assumption since a decrease in water density will cause keff to decreasa. Boiling is not permitted to occur under any circumstances. The design basis for the wet fuel storage criticality analysis is that there is a 95% confidence level that the keff of the fuel storage array will be less than 0.95 per ANS1 Standard N18.2-1973.
The results of the analysis for an Lifinite array of 17 x 17 assemblies en-riched to 3.5 w/o U-235 show that a 14.0 inch center-to-center rack spacing corresponds to at least 95% of the time keff will not exceed 0.95 at a 95% confidence level.
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The spent fuel storage racks have adequate energy absorption capabilities to withstand the impact of a dropped fuel assembly from the maximum lift height of the fuel handling bridge crane.
i In the analysis for both the new fuel storage racks and the spent fuel storage racks, the fuel assemblies
.are_ assumed to be in their most. reactive condition, namely fresh or undepleted and with no control rods
- or removable neutron absorbers present. Assemblies can not be closer together than the-design separation provided by the storage racks. 1Rua mechanical integrity of the fuel assembly is assumed. Verification that the design criteria for fuel storage are met is achieved through the use of standard Westinghouse Electric Corp-cration design methods such as the LEOPARD and PDQ codes.
Comment (4):
Confirm all degrees of credible interspersed moderation (e.g., from sprinklers) have been considered in the nuclear criticality safety evaluation.
Response
Nuclear criticality safety evaluations for-fuel stored in 4
21 inch center-to-center racks were performed assuming fuel ~
of the highest anticipated enrichment in place and optimum moderation conditions (such as foam or water mist arising from fire fighting operations) existing under dry storage conditions. The results of the evaluations showed that keff does not exceed 0.98.
Optimummoderationconditionsfrominterspersegmoderators require a density.of approximately 0.1 gram /cm.
Achieving moderator densities in this range is not credible. As a pointofreferenge,thewaterdensityofaheavyrainstorm is 0.00014 gm/cm ; steam at 1 ATM and 212 F has a density 3
of 0.0006'gm/cm ; firefighting sprinklers, foams, and sprays have densities of 0.001 gm/cm ; a stream from a water hose thatgasdivergedfrom1"to10"hasadensityof0.01 3
Thus, achieving a density of 0.1 gm/cm over a gm/cm.
significant rack volume has an extremely low probability of occurrence. The levels of low-density moderation needed to compromise the safety of the. fuel storage array are not achievable by accident.
2-Comment (5):
Provide the administrative controls to ensure " checkerboard" loading of new fuel in the spent fuel pool racks. What is to prevent the inadvertent insertion of an assembly in the open storage cells of the array?
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Response
A loading pattern will be developed for the storage of new fuel in the spent fuel pool. The loading pattern will' identify each fuel assembly's assigned storage location and will arrange the fuel in an
" expanded checkerboard" array.- This " expanded checkerboard" array is described in more detail in the response to Comment (6) below. Periodic loading verification checks will be performed during new fuel storage. operations in the sient fuel pool with a final verification to be performed af ter all new fuel assemblies have been loaded into their assigned storage locations.
The inadvertent lusertion of a new fuel assembly into one of the dry open storage cells of the expanded checker-board" array will have no adverse consequences since keff for this dry etorage condition will remain less than 0.98.
Only for the most optimum moderation conditions (i.e., fire fighting operations) could the 1:ef f exceed 0.98.
If during any of the periodic loading checks a fuel assembly is found to be out of its assigned location, it will be promptly returned to its correct storage location.
Comment (6):
Provide justification for the assumption the " checker-board" array of fuel assemblies has a multiplication factor (keff)'no greater than thnt for the array of assemblies in the new fuel storage areas. The edge-to l
edge spacing between some assemblies in the " checkerboard" array are closer than that between those in the new fuel storage racks.
Consideration should be given to the optimum credible water moderation within and between fuel assemblies.
Response
Dry storage of new fuel assemblies in the spent fuel pool t acks will be in an " expanded checkerboard" array such that an open storage cell exists in the 8 adjacent cells surround-i ing each assembly. Therefore, no two assemblies will be closer than the 21 inch center-to-center spacing ~of the new fuel storage racks. This more conservative loading pattern for new fuel storage in the spent fuel racks will result in a 32 inch center-to-center spacing between fuel assemblies.
For the flooded condition with unborated water present and assuming new fuel of the highest anticipated enrichment in place, the effective multiplication factor (keff) does not exceed 0.95.
For the dry storage condition, kef f does not exceed 0.98 with fuel of the highest anticipated. enrichment in place assuming possible sources of optimum moderation arising from fire fighting operations. The plastic covering around each assembly will be opened at the bottom to allow
^
water drainage should flooding and then draining of the fuel storage occur, thereby eliminating any possibility of.non-uniform radial moderator dist yh,ution.
Inadvertent insertion of an assembly in an open cell of. the array will be precluded
_ 8,
by the use of previously developed loading patterns and periodic loading verification checks.
(see Response to Comment 5 above) 9
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CPSES Fuel Building Floor Plan 1.
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Dear Mr. Ketzlach:
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As per your request, the additional information needed to complete your-review of our application for authorization for receipt., possession, storage, inspection and' preparation for transport of special nuclear material required for the operation of the-Comanche Peak Steam Electric Station Unit 1 is enclosed as page changes to the original application dated July 10, 1980. This enclosure includes the information previously-transmitted by-my-letter of April 3, 1981 and also the responses to your comments of May 21, 1981. Eight (8) copies of the responses to your comments are enclosed for your evaluation.
Your prompt completion of ~the application review will be appreciated.
Respectfully, R.
. Gary BWC:mac Enclosures g-m,. q*;,; V.i%j I
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Change Number 1 to Texas Utilities Generating Company Comanche Peak Steam Electric Station Unit l-
-Application for'Special Nuclear Material License Pages to be Removed New Pages to be Inserted Page Number Date Page Number Date 1
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7/16/81 i
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7/16/81 2A 7/16/81.
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TEXAS UTILITIES GENERATING COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 APPLICATION FOR SPECIAL NUCLEAR MATERIAL LICENSE Tnis application is filed pursuant to Title 10, Chapter 1, Code of Federal Regulations, Part 70 for authorization to receive, possess, store, inspect, and package for transport unirradiated nuclear fuel assemblies for Unit 1 of the Comanche Peak Steam Electric Station (CPSES). The term of the Special Nuclear Material License requested is for the period beginning October 1,1981 until receipt of the permanent l1 operating license.
The applicants are Texas Utilities Generating Company (TUGCO), Dallas
' Power & Light Company (DPL), Texas Electric Service Company (TESCO),
Texas Power & Light Company (TPL), Texas Municipal Power Agency (TMPA) and Brazos Electric Power Cooperative, Inc. (BEPC). DPL, TESCO, TPL, TMPA, and BEPC (collectively the "0wners") respectively own 181/3%,
35 5/6%, 35 5/6%, 6 1/5%, 3 4/5% interest in the station as tenants in common. Neither TUGC0 nor the Owners are owned or controlled by an alien, foreign corporation, or foreign government.
TUGC0 is the lead Applicant, and as such acts as agent for the Owners for the design, construction, and operation, as well as in licensing matters, but will have no ownership interest.
The location of the offices and principal officers for TUGC0 and the Owners can be found in the Application of TUGC0 and Owners, Docket Nos.
50-445 and 50-446, for Operating Licenses (Class 103) for the Comanche Peak Steam Electric Station Units 1 and 2.
Communications p Jrsuant to this license application should be sent to:
Mr. R. J. Gary Executive Vice President and General Manager Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201
)
l 1
, JULY 16, 1981 un
1.0 GENERAL INFORMATION 1.1 Reactor and Fuel 1.1.1 General The Comanche Peak Steam Electric Station is located in Somervell County, Texas, 65 air miles southwest of the Dallas - Fort Worth Metropolitan Area in North Central Texas. A detailed description of the geographic location is provided in the CPSES Final Safety Analysis Report (FSAR) Section 2.1.1.1, Docket Nos. 50-445 and 50-446. The Unit I construction permit number is CPPR-126 and the Unit 1 Reporting Identification Symbol as assigned by the Nuclear Regulatory Commission is YGL.
1.1.2 Fuel Assemblies The nuclear fuel assemblies consist of slightly enriched uranium dioxide pellets encased in Zircaloy-4 rods. The average outside diameter of the slightly enriched uraniun dioxide pellets is 0.3225 inches. The i
fuel rod pitch in a fuel assembly is 0.496 inches. The zircaloy fuel rods have a nominal outside diameter and wall thickness of 0.374 inches and 0.0225 inches, respectively.
Each assembly contains 264 fuel rods, 24 Zircaloy-4 control rod guide thimbles, and 1 Zircaloy-4 instrumentation thimble arranged in a 17 x 17 matrix.
The 17 x 17 matrix is maintained by 8 inconel grid assemblies located along the length of the fuel assembly. The assembly top and bottom nozzles are constructed of stainless steel. The assembly is approximately 160 inches in length with a nominal active fuel length of 144 inches.
Each assembly is approximately 8.4 inches square.
1.1.3 Assembly Enrichment and Weights The initial core contains nominal assembly enrichmeate of 1.60 w/o, 2.40 w/o, and 3.1 w/o U-235. The tot uranium weight per a'ssembly is nominally 461 Kg of which less than 15 Kg is contained as U-235. The total assemby design weight, including structural components, is 665 Kg. The fuel assemblies contain no U-233, Pu, depleted uranium, or thorium.
Disregarding the nominal enrichment variations produced during the U.S. Department of Energy's enriching process, the maximum fuel assembly enrichment to be i
stored under the Comanche Peak Steam Electric Station Unit 1 Special Nuclear Material License is 3.1 weight JULY 16, 1981 _. _ _
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1 x
percent (w/o) U-235. The highest anticipated enrichment assumed for nuclear criticality safety 1
analyses is 3.5 w/o U-235.
1.1.4 Total Fuel Assemblies and Uranium The total number of fuel assemblies in the initial core
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is 193. The total weights of U-235 and uranium are approximately 2,100 Kg and 89,060 Kg, respectively.
1 4
JULY 16, 1981
- 2A-
1.2 Storage Conditions 1.2.1 Fuel Storage Area Fuel storage and handling operations will be perfomed in the Fuel Handling Building. One hundred thirty-two (132) fuel assemblies will be stored in the new fuel storage area. The remaining 61 fuel as:;emblies for the initial core will be placed in storage in the spent fuel pool storage racks. The new fuel storage area capacity can be increased to 140 assemblies. Should this upgrading occur, the remaining 53 fuel assemblies will be placed in storage in the spent fuel pool storage racks. Detailed elevation and plan views of the Fuel Handling Building are shown in FSAR Figures 1.2-38 through 1.2-40.
Temporary storage of new fuel assemblies in their shipping containers may be necessary for short periods of time during new fuel receipt.
If such storage is required, the new fuel will be stored in a horizontal position in a closed shipping container. The container will be stored on the transportation vehicle or in the new fuel receipt area (Fuel Building elevation 841').
The storage location of any new fuel assembly in the new fuel receipt area will be no closer than twelve (12) feet from the nearest new fuel inspection station.
If more than one new fuel inspection station is etablished, the minimtm distance between the inspection stations will be twelve (12) feet. The minimum spacing 1
between the nearest fuel storage racks (i.e., the new fuel storage racks) and the new fuel receipt and inspection area is a horizontal distance of twelve (12) feet.
It should be noted, however, that the new fuel storage racks are located in a different area of the i
Fuel Building and are separated from the new fuel rece"t and inspection area by a reinforced concrete wall. The new fuel receipt and inspection operating area is located nineteen (19) feet below (Fuel Building elevation 841') the new fuel racks operating area (Fuel Building elevation 860').
Figure 1 shows the Fuel Building location of the fuel storage areas and the new fuel receipt and inspection area.
1.2.2 Fuel Storage Area Activities Only those activities which involve new fuel receipt, fuel inspection, and fuel handling and storage are normally conducted in or adjacent to the fuel handling and storage areas. No construction activities which could possibly result in damage to the fuel will be JULY 16, 1981 [
allowed in the fuel storage areas during fuel handling, inspection or storage.
1.2.3 Fuel Handling Building Equipment and Systems The Fuel Handling Building structures, components, equipment, systems, and the design criteria used to assure their structural integrity are described in FSAR Sections 9.1 and 3.8.4.
Prior to receipt of new fuel, all required fuel handling equipment and storage facilities will be inspected and tested to ensure safe operation during 3
fuel handling activities.
I 1.2.4 Fire Alam ind Control Systems As presented in FSAR Figure 9.5-40, the Fuel Handling Building combustible loading classification is low based on tables from the National Fire Protection Association (NFPA) and Copper Life Safety Fire Sprinkler System Handbooks. The barriers separating the fire areas are constructed of concrete block or poured, reinforced concrete, or both with approved fire doors, fire dampers, and penetrations of an equivalent rating.
Fire protection will be provided by portable extinguishers, hose stations, and a remote manual deluge system. Fire detection is provided by ionization and flame detectors equipped with both local and remote alarms. A detailed analysis of the fire protection plans is discussed in FSAR Section 9.5.1.3.5.
APCSB 9.5-1, Appendix A, requires that the fire protection program (plans, personnel and equipment) for buildings storing new reactor fuel and for adjacent fire zones which could affeu the feel storage zone to be fully operational before fuel is received at the site. Therefore, the Fire Protection Programs for the 1
Fuel Handling Building will be in effect and the suppression systems operational prior to receipt of fuel on the site as stated in the CPSt.S FSAR Section 9.5.
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2.0 HEALTH AND SAFETY 2.1 Radiation Control 2.1.1 Training and Experience The training and experience of the Comanche Peak Steam Electric Station Health Physics personnel is described l1 in FSAR Section 13.1.3.2.
2.1.2 Procedures and Equipment Administrative controls which govern the safe handling and storage of fuel will be the responsibility of the Engineering Superintendent. Those procedures which control the safe handling of fuel will be approved by the Station Operations Review Committee (50RC). The function of the SORC is described in Section 13.4.1 of the CPSES FSAR.
1 The manipulation of the new fuel assemblies will be perfomed by CPSES Operations personnel trained in proper fuel handling techniques and, in addition, will be done in accordance with approved written fuel handling procedures containing provisions to assure that all fuel assemblies are handled correctly.
Radiation and contamination monitoring will be performed prior to the initial handling and storage of 1
new fuel. All new fuel that has not been urdoaded or unpacked will be handled as contaminated material with all appropriate radiological controls in effect until contamination checks are performed. New fuel will he checked for radioactive contamination by CPSES Radiation Protection personnel as part cf the new fuel l1 inspection procedure. Swipes or smears will be taken of the fuel in order to obtain a representative sample of the surface contamination of the entire assembly and t
will be counted for alpha and beta / gamma activity to determine the amount of contamination present.
If the l
amount of contamination is found to exceed allowable l
limits, the source of the contamination will be determined and appropriate decontamination steps will be initiated as required. This practice should identify any possible radiation hazards associated with external contamination of new fuel assemblies and allow 1
proper planning for ALARA controls deemed necessary Dy,
Radiation Protection personnel. The CPSES Health Physics Program outlined in FSAR Section 12.5 describes the procedures and equipment involved in radiological controls. JULY 16, 1981
Due to the fact that the fuel will be unirradiated, there will be no significant radiation hazard associated with the low level radioactivity of the fuel i
itself. The handling and storage of the fuel, as outlined above, will be sufficient to maintain radiation exposures ALARA.
2.1.3 Detection Calibration and Testing Testing of the detectors used to measure radioactive contamination on new fuel assemblies will consist of daily checks as required on background radiation, detector efficiency, and the updating of a daily trend plot of detector perfomance.
In addition, the instrumentation will be calibrated as a minimum on a quarterly basis using appropriate calibration sources.
2.2 Nuclear Criticality Safety After receiving the shipping containers at the plant site, only one metal shipping container with fuel assemblies will be opened at any one time. Each fuel assembly will be removed from its shipping container and inspected at a new fuel inspection station.
If more than one new fuel inspection station is established, the minimum distance between the inspection stations will be twelve (12) feet. Also, the distance between any new fuel inspection station and new fuel temporarily stored in the new fuel receipt area will be greater than er equal to twelve (12) feet. These separation requirements apply to new fuel inspection stations located in 1
the new fuel receipt area (Fuel Building elevation 841').
However, if an inspection station is to be located at Fuel Building elevation 860' near the new and spent fuel storage areas, a minimum separation of twelve (12) feet will also be maintained between the inspection station and the new and spent fuel storage racks.
If no defects are found, the fuel assembly will be moved to the fuel storage racks. The fuel i
storage racks in the new fuel storage area and spent
-5A-JULY 16, 1981
fuel pool are designed for a nominal 21 inch and 16 inch, respectively, center-to-center spacing between fuel assembly storage cells.
The new fuel storage racks (Figure 2) are composed of individual vertical cells fastened together to fom a module which can be fimly bolted to anchors in the floor of the new fuel storage pit. The new fuel storage racks are designed to include storage for two-thirds core at a center-to-center spacing of 21 inches. The design of the fuel storage rack assembly is such that it is impossible to insert the new fuel assemblies in other than prescribed locations. A metal cover will be positioned over each new fuel storage rack section after each section is loaded. All surfaces that come into contact with the fuel assemblies are made of annealed austenitic stainless steel. The fuel storage racks are designed to withstand normal operating loads as well as Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (0BE) seismic loads meeting ANS Safety Class 3 and ASME B&PV Code Section 111, Appendix XVII requirements. The fuel storage racks are also designed to meet the seismic Category I requirements of NRC Reguletory Guide 1.29, Revision 2, February 1976. The fuel storage racks can withstand an uplift force equal to the 5000 lb. uplift force of the fuel handling bridge crane.
The spent fuel storage racks (Figure 3) are composed of 1
individual vertical cells fastened together at a 16 inch center-to-center spacing to form a module which is firmly bolted to anchors in the floor of the spent fuel pool.
Space between storage cells is blocked to prevent insertion of fuel.
All surfaces that come into contact with fuel assemblies are made of annealed austenitic stainless steel. The spent fuel storage racks are designed to withstand shipping, handling, nomal operating loads (dead loads of fuel assemblies), as well as SSE loads. These racks meet ANS Safety Class 3 and ASME B&PV Code,Section III, Appendix UII requirements. The spent fuel storage racks are also designed to meet the seismic Category I requirements of Regulatory Guide 1.29, Revision 2, February 1976. The racks can withstand an uplift l
force equal to the uplift force of the spent fuel pool bridge l
hoist.
Both the new and spent fuel storage racks have adequate energy l
absorption capabilities to withstand the impact of a dropped fuel assembly from the maximum lift height of the fuel handling bridge crane. An analysis was done using a standard 17 x 17 fuel assembly with the handling tool and a total mass of 2000 lb. falling from a height of 3.5 feet (without dampng or energy dissipation) on to the top of a fuel cell.
The analysis results show tnat the fuel cell deforms in compression and shortens in length. The accident would not JULY 16, 1981 -
result ir. an unsafe geometric spacing of fuel assemblies.
Cranes carrying loads heavier than a fuel assembly and its associated handling tool will be prevented by interlocks or administrative controls, or both, from traveling over the new fuel storge area when new fuel is stored in the new fuel storage racks.
New fuel is stored in.21 inch center-to-center racks in the new fuel storage facility with no water present. The plastic covering around each assembly will be opened at the bottom to allow water drainage should flooding and then drainage of the fuel storage area occur. These racks are designed to prevent accidental criticality even if unborated water is present.
The design of the new fuel storage racks is such that the effective multiplication factor (keff) does not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry or flooded conditions).
For the normally dry condition, keff does not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming possible sources of moderation such as those that could arise dcring fire fighting operations (such as foam or water mist).
Consideration is given to the inherent neutron absorbing effect of the materials of construction.
Nuclear criticality safety evaluations for fuel stored in 21 i
inch center-to-center racks were performed assuming fuel of the highest anticipated enrichment in place and opti:num moderation conditioas (such as foam or water mist arising from fire fighting operations) existing under dry storage conditions. Optimum moderation conditions from interspersed moderators require a density of approximately 0.1 gram /cm3 Achieving moderator densities in this range is not credible.
As a point of reference, the water density of a heavy 0.00014gmfcm3;steamat1ATMand2120Fhasa rainstorm is density of 0.0006 gm/cm ; fitefighting sprinklers, foams, and sprays have densities of 0.002 gm/cm3; a stream from a water I
hose that has diverged from 1" to 10" has a dengity of 0.01 gn/cmd. Thus, achieving a density of 0.1 gm/cm3 over a significant rack volume has an extremely low probability of occurrence. The levels of low-density moderation needed to compromise the safety of the fuel storage array are not achievable by accident. The results of the evaluations showed that keff does not exceed 0.98.
Dry storage of new fuel assemblies in the spent fuel pool-racks will be in an " expanded checkerboard" array such that an open storage cell exists in the 8 adjacent cells surroending each assembly. Therefore, no two assemblies will be closer than the 21 inch center-to-center spacing of the new fuel storage racks. This more conservative loading pattern for new fuel storage in the spent fuel racks will result in a 32 inch center-to-center spacing between fuel assemblies. The plastic
-6A-JULY 16, 1981
covering around each assembly will be opened at the bottom to allow water drainage should flooding and then draining of the fuel storage area occur, thereby elimir;ating any possibility of non-unifom radial moderator distributions.
Inadvertent insertion of an assembly in an open cell of the array will be precluded by tha use of previously developed loading patterns and loading verification checks.
A loading pattern will be developed for the storage of new fuel in the spent fuel pool. The loading pattern will identify each fuel assembly's assigned storage location and will arrange the fuel in an " expanded checkerboard" array.
The individual conducting new fuel loading into the spent fuel pool will verify correct assembly location after insertion of each new fu-l assembly into its assigned storage rack. An independent loading verification will also be conducted after each assembly insertion by a second individual participating in fuel storage operations.
In addition, a loading check will be conducted by CPSES Reactor Engineering after each shipment of fuel is off-loaded in assigned storage locations. A
" shipment" of new fuel will consist of ro more than twelve (12) fuel assemblies.
If during any of the loading checks a 1
fuel assembly is found to be out of its assigned location it will be promptly returned to its correct storage location.
The inadvertent insertion of a new fuel assembly into one of the dry open storage cells of the " expanded checkerboard" array will have no adverse consequences since keff for this dry storage condition will remain less than 0.98.
Notwithstanding the conservative " expanded checkerboard" loading pattern in the spent fuel pool racks, a criticality analysis was performed assuming unborated water of 1.0 gm/cm3 and new fuel of the highest anticipated enrichment in place.
Overgherangeofwgterdensitiesofinterest(corresponding to 60 F through 212 F), full density water is a conservative assumption since a decrease in water density will cause keff to decrease. Boiling is not pemitted to occur under any circumstances.
The design basis for the wet fuel storage criticality analysis is that there is a 95% confidence level that the keff of the fuel storage array will be less than 0.95 per ANSI Standard N18.2-1973. The results of the analysis for an infinite array of 17 x 17 assemblies enriched to 3.5 w/o U-235 show that a 14.0 inch center-to-center rack spacing l
l corresponds to at least 95% of the time keff will not exceed l
0.95 at a 95% confidence level.
JULY 16, 1981
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Figure 2.
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Figure 3.
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1 In the analysis for both the new fuel storage racks and the spent fuel storage racks, the fuel assemblies are assumed to be in their most reactive condition, namely fresh or undepleted and with no control rods or removable neutron absorbers present. Assemblies can not be closer together than the design separation provided by the storage racks. The mechanical integrity of the fuel assembly is assumed.
Verification that the design criteria for fuel storage are met 1
is achieved through the use of standard Westinghouse Electric Corporation design methods such as the LEOPARD and PDQ codes.
Detailed explanations of the criticality safety studies and their results are presented in CPSES FSAR Sections 9.1 and 4.3.2.6.
New fuel elements will be removed from their usual storage locations from time to time for such activities as fuel assembly relocation in storage and fuel inspection. The manipulation of the new fuel assemblies will be performed by CPSES operations personnel trained in proper fuel handling techniques and, in addition, will use fuel handling procedures which contain provisions to assure that fuel assemblies are handled correctly. Equipment and structures used for fuel handling activities are designed to provide for safe operation as described in FSAR Section 9.1.4.
In order to prevent accidental nuclear criticality, only one new fuel assembly will be allowed to be removed from'a shipping container or an approved storage location at any one I
time. Further discussion of the criticality of fuel assemblies is found in FSAR Section 4.3.2.6.
Because of the fuel storage facilities design and administrative controls limiting the maximum number of fuel assemblies allowed out of the storage locations, the possibility of accidental criticality during receb+.,
inspection, and other handling activities is eliminated.
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Therefore, an exemption in whole from the requirements of 10 CFR 70.24 is requested as provided by 10 CFR 70.24(d).
l 2.3 Accident Analysis Interlocks or administrative controls, or both, prevent the Fuel Building handling equipment capable of carrying loads heavier than a fuel assembly from traveling over the fuel storage area. The fuel storage racks are designed to maintain a safe geometric spacing of fuel assemblies despite the impact l
of a fuel assembly dropped from the maximtrn lift height of the i
fuel handling bridge crane.
l JULY 16, 1981 i l
Since the license will involve only the handling and storage of unirradiated reactor fuel, there would be no significant safety hazard as a tasult of a fuel handling accident-due to the absence of any fission products in the fuel hanaling areas. Should a fuel handling accident result in the release of any of the uranium contents of the new fuel, JULY 16, 1981
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