ML20010B548
| ML20010B548 | |
| Person / Time | |
|---|---|
| Site: | 07002932 |
| Issue date: | 07/16/1981 |
| From: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20010B544 | List: |
| References | |
| 19412, NUDOCS 8108170203 | |
| Download: ML20010B548 (16) | |
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Change Number 1 to Texas Utilities Generating Company Comanche Peak Steam Electric Station Unit 1 Appl' cation for Special Nuclear Material License Pages to be Removed New Pages to be Inserted Page Number Date Page Number Date 4
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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 tenn of the Special Nuclear Material License requested is for the period beginning October 1,1981 until rei af pt of the pennanent i
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 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 Tower Dallas, Texas 75201 JULY 16, 1981
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4 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 rinal 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 uranium 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 a ssembly. 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 nominally 461 Kg of which less than 15 Kg is contained as U-235. The total assembly 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 Statio..
Unit 1 Special Nuclear Material License is 3.1 weight JULY 16. 1981 _
percent (w/o)U-235. The highest anticipated enrichment assumed for nuclear criticality safety 1
analyses is 3.5 w/o U-235.
i 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, i
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1.2 Storage Conditions 1.2.1 Fuel Storage Area Fuel storage and handling operations will be perfonned in the Fuel Handling Building. One hundred thirty-two (132) fuel assemblie;, 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 assemblies will be placed in storage in the spent fuel pool storage racks. Detailed elevation and plan views of the Fuel Handling Nilding are shown in FSAR Figures 1.2-38 through 1.2-40.
Temporary storage of new feel 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').
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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 minimtsn distance between the inspection stations will be twelve (12) feet. The minimum spacing I
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 Fuel Building and are separated from the new fuel receipt 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 i shows the Fuel Building location of the fuel storage areas ano 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 nonnally 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
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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 1
inspected and tested to ensure safe operation during fuel handling activities.
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 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 affect the fuel storage zone to be fully operational before fuel is received at the site. Therefore, the Fire Protection Programs for the i
Fuel Hardling Building will be in effect and the suppression systems operational prior to receipt of i
fuel on the site as stated in tho CPSES FSAR Section 9.5.
JULY 16, 1981
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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 (SORC). 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 performed 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 unloaded or unpacked will be handled as contaminated material with all appropriate radiological controls in effect until contamination checks are performed. New fuel will be checked for radioactive contamination by CPSES Radiation Protection personnel as part of the new fue!
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 will be counted for alpha and beta / gamma activity to determine the amount of contamination present.
If the amount cf contamination is found to exceed allowabli 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 by 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 1
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 performance.
In addition, the instramentation 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 L0 tween any new fuel inspection station and new fuel temporarily stored in the new fuel receipt area will be greater than or 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 tweb.e (12) feet will also be maintained between the inspectirn 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 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 incividual vertical cells fastened together to fom a module which can be firmly 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 spa;ing 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 a-e designed to withstand nomal 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 fuel storage racks are also designed to meet the seismic Category I requirements of NRC RegJlatory 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 fom 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 MME 'i&PV Code,Section III, Appendix XVII requirements. The speat fuel storage racks are also designed to meet the seisnde Category I requirements of Regulatory Guide 1.29, Revision 2, February 1976. The racks can withstand an uplift force equal to the uplift force of the spent fuel pool bridge hoist.
Both the new and spent fuel storage racks have adequate energy absorption capabilities to withstand the impact of a dropped fuel assembly from the maximum lift height oi the fuel handling bridge crane. An analysis was done using a standard 17 x 17 fuel assembly with the handling tool ar.d a total mass of 2000 lb. falling from a height of 3.5 feet (without damping or energy dissipation) in to the top of a fuel cell.
The analysis results show that the fuel cell defoms in compression and shortens in length. The accident would not JULY 16,1981 -
result in 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, frorn traveling over the new fuel stcrge 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 bniichment in place, assuming optimum moderation (under dry or flooded conditions).
For the nomally 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 during fire fighting operations (such as foam or water mist).
Consi'teration is given to the inherent neatron 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 optimum moderation conditions (such as foam or water mist arising from fire fighting operations) existing under dry storage conditions. Optimum moderation conditions from intersperged moderators rer, dire 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 rainstorm is 0.00014 gm/cm3; steam at 1 ATM and 2120F has a density of 0.0006 gm/cm4; firefighting sprinklers, foams, and sprays have densities of 0.001 gm/cm3; a stream from a water hose that has diverged from 1" to 10" has a dengity of 0.01 gm/cm3 Thus, achieving a density of 0.1 gm/cmo 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 r.ew 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 surrounding 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 t
covering around each assa:.bly will be opened at tne bottom to allow water drainage should flooding anc then d. 3fning of the fuel storage area occur, thereby eliminating any possibility of non-unifom radial moderator Jistributions.
Inadvertent insertion of an assembly in an open cell of the array will be precluded by the use of previously developed loading patterns and loading verification checks.
A loading pattern will be developed 'or the storage of new fuel in the spent fuel pool.
The 1 Jading pattern will identify each fuel assembly's assigned storage location and will arrange the fuel in an '; expanded checkerboard" array.
The individual ccnducting new fuel loading into the spent fuel pool will verify correct assembly location after insertion of each new fuel 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 no more than twelve (12) fuel assemblies.
If during any of the loading checks a 1
fuel assev 'y 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 conservati"e " 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.
Overgher?ngeofwgterdensitiesofinterest(corresponding to 60 F id cagh 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 sho,i 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.
JULY 16, 1981
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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 conditic.7, 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,JQ 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 1
shipping container or an approved storage location at any one time. Further discussion of the criticality of fuel cssemblias 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 receipt, l
inspection, and other handling activities 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 l
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 anemblies despite the impact of a fuel assembly dropped from the maximum lift height of the fuel handling bridge crane.
I JL'LY 16,1981 _ _
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, I
JULY 16, 1981 1911'3